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Sample records for neutron-irradiated titanium-modified stainless

  1. Irradiation creep and creep rupture of titanium-modified austenitic stainless steels and their dependence on cold work level

    International Nuclear Information System (INIS)

    Garner, F.A.; Hamilton, M.L.; Eiholzer, C.R.; Toloczko, M.B.; Kumar, A.S.

    1991-11-01

    A titanium-modified austenitic type stainless steel was tested at three cold work levels to determine its creep and creep rupture properties under both thermal aging and neutron irradiation conditions. Both the thermal and irradiation creep behavior exhibit a complex non-monotonic relationship with cold work level that reflects the competition between a number of stress-sensitive and temperature-dependent microstructural processes. Increasing the degree of cold work to 30% from the conventional 20% level was detrimental to its performance, especially for applications above 550 degrees c. The 20% cold work level is preferable to the 10% level, in terms of both in-reactor creep rupture response and initial strength

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

  3. Microstructural response of titanium-modified austenitic stainless steels to neutron exposure of 70 dpa in FFTF/MOTA

    International Nuclear Information System (INIS)

    Katoh, Yutai; Kohno, Yutaka; Kohyama, Akira

    1994-01-01

    JPCA, a titanium-modified austenitic stainless steel, in solution-annealed or cold-worked condition and a compositionally modified JPCA in solution-annealed condition were examined by transmission electron microscopy following irradiation in FFTF/MOTA to an exposure level of up to about 70 dpa at 390 to 600 C. At lower temperatures, all the materials developed qualitatively similar cavity-, dislocation- and precipitate-microstructures. The lower-temperature swelling peak, which appeared at near 410 C, was more efficiently suppressed by phosphorus addition than cold-working. Irradiation at or above 520 C produced substantially large swelling in solution-annealed JPCA. The cavities contributed to this higher-temperature swelling developed in association with M 6 C-type precipitates. Neither cavities other than very small helium bubbles nor massive particles of M 6 C-type precipitates were observed in cold-worked and phosphorus-modified materials, in which MC-type precipitates developed at very high concentration. The effect of pre-irradiation microstructure and compositional modification on the behavior of these precipitates is discussed. ((orig.))

  4. Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1993-01-01

    Austenitic stainless steels are important structural materials common to several different reactor systems, including light water and fast breeder fission, and magnetic fusion reactors (LWR, FBR, and MFR, respectively). The microstructures that develop in 300 series austenitic stainless steels during neutron irradiation at 60-700 C include combinations of dislocation loops and networks, bubbles and voids, and various kinds of precipitate phases (radiation-induced, or -enhanced or -modified thermal phases). Many property changes in these steels during neutron irradiation are directly or indirectly related to radiation-induced microstructural evolution. Even more important is the fact that radiation-resistance of such steels during either FBR or MFR irradiation is directly related to control of the evolving microstructure during such irradiation. The purpose of this paper is to provide an overview of the large and complex body of data accumulated from various fission reactor irradiation experiments conducted over the many years of research on microstructural evolution in this family of steels. The data can be organized into several different temperature regimes which then define the nature of the dominant microstructural components and their sensitivities to irradiation parameters (dose, helium/dpa ratio, dose rate) or metallurgical variables (alloy composition, pretreatment). The emphasis in this paper will be on the underlying mechanisms driving the microstructure to evolve during irradiation or those enabling microstructural stability related to radiation resistance. (orig.)

  5. Study of irradiation damage structures in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, Shozo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-08-01

    The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs.

  6. Study of irradiation damage structures in austenitic stainless steels

    International Nuclear Information System (INIS)

    Hamada, Shozo

    1997-08-01

    The irradiation damage microstructures in austenitic stainless steels, which have been proposed to be a candidate of structural materials of a fusion reactor, under ions and neutrons irradiation have been studied. In ion irradiation experiments, cross-sectional observation of the depth distribution of damage formed due to ion irradiation became available. Comparison and discussion between experimental results with TEM and the calculated ones in the depth profiles of irradiation damage microstructures. Further, dual-phase stainless steels, consisted of ferritic/austenitic phases, showed irradiation-induced/enhanced precipitation during ion irradiation. High Flux Isotope Reactor with high neutron fluxes was employed in neutron-irradiation experiments. Swelling of 316 steel showed irradiation temperature dependence and this had strong correlation with phase instability under heavy damage level. Swelling resistance of Ti-modified austenitic stainless steel, which has good swelling resistance, decreased during high damage level. This might be caused by the instability of Ti-carbide particles. The preparation method to reduce higher radioactivity of neutron-irradiated TEM specimen was developed. (author). 176 refs

  7. 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)

  8. Swelling behavior of titanium-modified AISI 316 alloys

    International Nuclear Information System (INIS)

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

    1984-01-01

    It appears that titanium additions to stainless steels covering a wide compositional range around the specifications of AISI 316 result only in an increased delay period before neutron-induced void swelling proceeds. Once swelling is initiated the post transient behavior of both annealed and cold-worked titanium-modified steels is quite consistent with that of AISI 316, approaching a relatively temperature-independent swelling rate of approx. 1% per dpa

  9. Evaluation of neutron irradiation effect on SCC crack growth behaviour for austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    Austenitic stainless steels are widely used as structural components in reactor pressure vessel internals because of their high strength, ductility, and fracture toughness. However, exposure to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the steels. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term plant operation in high temperature water environments is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) in Type 304 stainless steel in BWR environment. JNES had been conducting IASCC project during the JFY (2000) - JFY (2008) period, and prepared an engineering database on IASCC. However, the data of Crack Growth Rate (CGR) below the critical fluence level are not sufficient. So, this project was initiated to obtain the CGR data below the critical fluence level. Test specimens have been irradiated in the Halden reactor, operating by the OECD Halden Reactor Project, and the post irradiation examination (PIE) will be conducted from JFY (2011) to JFY (2013), finally the modified IASCC guide will be prepared in JFY (2013). (author)

  10. Austenitic stainless steel alloys having improved resistance to fast neutron-induced swelling

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Rowcliffe, A.F.; Leitnaker, J.M.

    1979-01-01

    The present invention is based on the discovery that radiation-induced voids which occur during fast neutron irradiation can be controlled by small but effective additions of titanium and silicon. The void-suppressing effect of these metals in combination is demonstrated and particularly apparent in austenitic stainless steels

  11. Weldability of neutron-irradiated stainless steel and nickel-base alloy

    International Nuclear Information System (INIS)

    Koyabu, Ken; Asano, Kyoichi; Takahashi, Hidenori; Sakamoto, Hiroshi; Kawano, Shohei; Nakamura, Tomomi; Hashimoto, Tsuneyuki; Koshiishi, Masato; Kato, Takahiko; Katsura, Ryoei; Nishimura, Seiji

    2000-01-01

    Degradation of of weldability caused by helium, which is generated by nuclear transmutation irradiated material, is an important issue to be addressed in planning of proactive maintenance of light water reactor core internal components. In this work, the weldability of neutron.irradiated stainless steel and nickel-base alloy, which are major constituting materials for components, was practically evaluated. The weldability was first examined by TIG welding in relation to the weld heat input and helium content using various specimens (made of SUS304 and SUS316L) sampled from reactor internal components. The specimens were neutron irradiated in a boiling water reactor to fluences from 4 x 10 24 to 1.4 x 10 26 n/ m 2 (E> l MeV ), and resulting helium generation ranged from 0.1 to 103 appm. The weld defects were characterized by dye penetrant test and cross-sectional metallography. The weldability of neutron-irradiated stainless steel was shown to be better at lower weld heat input and lower helium content. To evaluate mechanical properties of welded joints, thick plates (20 mm) specimens of SUS304 and Alloy 600 were prepared and irradiated in Japan Material Test Reactor (JMTR). The helium content of the specimens was controlled to range from 0.11 to 1.34 appm selected to determine threshold helium content to weld successfully. The welded joints had multiple passes by TIG welding process at 10 and 20 kJ/cm heat input. The welded joints of thick plate were characterized by dye penetrant test, cross-sectional metallography, tensile test, side bend test and root bend test. It was shown that irradiated stainless steel containing below 0.14 appm of helium could be welded with conventional TIG welding process (heat input below 20 kJ/cm). Nickel-base alloy, which contained as much helium as stainless steel could be welded successfully, could also be welded with conventional TIG welding process, These results served as basis to evaluate the applicability of repair welding to

  12. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  13. DT fusion neutron irradiation of BPNL niobium nickel and 316 stainless steel at 1750C

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation at 175 0 C of 17 niobium wires, one niobium foil, 14 316 stainless steel wires, one 316 stainless steel foil, nine nickel wires, and two nickel foils from BPNL is described. The sample position, beam-on time, neutron dose record, and neutron fluence are given

  14. Austenitic stainless steel alloys having improved resistance to fast neutron-induced swelling

    International Nuclear Information System (INIS)

    Bloom, E.E.; Stiegler, J.O.; Rowcliffe, A.F.; Leitnaker, J.M.

    1977-01-01

    The present invention is based on the discovery that radiation-induced voids which occur during fast neutron irradiation can be controlled by small but effective additions of titanium and silicon. The void-suppressing effect of these metals in combination is demonstrated and particularly apparent in austenitic stainless steels. 3 figures, 3 tables

  15. Effects of Si and Ti on the phase stability and swelling behavior of AISI 316 stainless steel

    International Nuclear Information System (INIS)

    Lee, E.H.; Rowcliffe, A.F.; Kenik, E.A.

    1979-01-01

    The swelling behavior of neutron irradiated stainless steels is strongly influenced by solute segregation and precipitation phenomena. The extent to which in-reactor swelling behavior may be simulated by heavy ion irradiation depends upon the extent to which in-reactor phase changes are reproduced; this question is addressed by comparing the precipitation behavior under neutron irradiation with behavior during 4 MeV Ni ion irradiation for AISI 316 stainless steel and a related stainless steel containing additions of titanium and silicon. The results are discussed qualitatively in terms of the effects of damage rate on solute segregation and the effects of displacement cascades on the dissolution of particles. It is shown that the partitioning of elements into various phases during irradiation is not a sufficient condition for the iniatiation of swelling in stainless steels modified with silicon and titanium. It is also necessary for helium to be generated simultaneously with the breakdown of the matrix into various phases; it is believed that helium trapping at the growing particle-matrix interface is responsible for the observed physical association between voids and precipitates. (Auth.)

  16. Effects of Si and Ti on the phase stability and swelling behavior of AISI 316 stainless steel

    International Nuclear Information System (INIS)

    Lee, E.H.; Rowcliffe, A.F.; Kenik, E.A.

    1978-01-01

    Swelling behavior of neutron irradiated stainless steels is influenced by solute segregation and preciptation phenomena. The extent to which in-reactor swelling behavior may be simulated by heavy ion irradiation depends upon the extent to which in-reactor phase changes are reproduced; this question is addressed by comparing the precipitation behavior under neutron irradiation with behavior during 4 MeV Ni ion irradiation for AISI 316 stainless steel and a related stainless steel containing additions of titanium and silicon. The results are discussed qualitatively in terms of the effects of damage rate on solute segregation and the effects of displacement cascades on the dissolution of particles. It is shown that the partitioning of elements into various phases during irradiation is not a sufficient condition for the initiation of swelling in stainless steels modified with silicon and titanium. It is also necessary for helium to be generated simultaneously with the breakdown of the matrix into various phases; it is believed that helium trapping at the growing particle-matrix interface is responsible for the observed physical association between voids and precipitates

  17. Dose dependence of the microstructural evolution in neutron-irradiated austenitic stainless steel

    International Nuclear Information System (INIS)

    Zinkle, S.J.; Maziasz, P.J.; Stoller, R.E.

    1993-01-01

    Microstructural data on the evolution of the dislocation loop, cavity, and precipitate populations in neutron-irradiated austenitic stainless steels are reviewed in order to estimate the displacement damage levels needed to achieve the 'steady state' condition. The microstructural data can be conveniently divided into two temperature regimes. In the low temperature regime (below about 200 degrees C) the microstructure of austenitic stainless steel is dominated by 'black spot' defect clusters and faulted interstitial dislocation loops. The dose needed to approach saturation of the loop and defect cluster densities is generally on the order of 1 displacement per atom (dpa) in this regime. In the high temperature regime (∼300 to 700 degrees C), cavities, precipitates, loops and network dislocations are all produced during irradiation; doses in excess of 10 dpa are generally required to approach a 'steady state' microstructural condition. Due to complex interactions between the various microstructural components that form during irradiation, a secondary transient regime is typically observed in commercial stainless steels during irradiation at elevated temperatures. This slowly evolving secondary transient may extend to damage levels in excess of 50 dpa in typical 300-series stainless steels, and to >100 dpa in radiation-resistant developmental steels. The detailed evolution of any given microstructural component in the high-temperature regime is sensitive to slight variations in numerous experimental variables, including heat-to-heat composition changes and neutron spectrum

  18. Evaluation of neutron irradiation effect on SCC crack growth behaviour of austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    Austenitic stainless steels are widely used as structural materials alloy in reactor pressure vessel internal components because of their high strength, ductility and fracture toughness. However, exposure due to neutron irradiation results in changes in microstructure, mechanical properties and microchemistry of the material. Irradiation assisted stress corrosion cracking (IASCC) caused by the effect of neutron irradiation during long term operation in high temperature water environments in nuclear power plants is considered to take the form of intergranular stress corrosion cracking (IGSCC) and the critical fluence level has been reported to be about 5x10{sup 24}n/m{sup 2} (E>1MeV) for Type 304 SS in BWR environment. JNES had been conducting IASCC project during from JFY 2000 to JFY 2008, and prepared an engineering database on IASCC. However, the data of crack growth rate (CGR) below the critical fluence level are not sufficient. Therefore, evaluation of neutron irradiation effect project (ENI) was initiated to obtain the CGR data below the critical fluence level, and prepare the SCC growth rate diagram for life time evaluation of core shroud. Test specimens have been irradiated in the OECD/Halden reactor, and the post irradiation experiments (PIE) have been conducting during from JFY 2011 to JFY 2013, finally the modified IASCC guide will be prepared in JFY 2013. (author)

  19. High Ni austenite stainless steel resistant to neutron irradiation degradation

    International Nuclear Information System (INIS)

    Yonezawa, Toshio; Iwamura, Toshihiko; Kanasaki, Hiroshi; Fujimoto, Koji; Nakata, Shizuo; Ajiki, Kazuhide; Nakamura, Mitsuhiro.

    1997-01-01

    The composition of the stainless steel of the present invention comprises from 0.005 to 0.08% of C, up to 3% of Mn, up to 0.2% of Si+P+S, from 25 to 40% of Ni, from 25 to 40% of Cr, up to 3% of Mo, up to 0.3% of Nb+Ta, up to 0.3% of Ti, up to 0.001% of B and the balance of Fe. A solid solubilization treatment at a temperature of from 1,000 to 1,150degC is applied to the stainless steel having the composition. The stainless steel is excellent in stress corrosion cracking-resistance at a working circumstance of a LWR type reactor (high temperature and high pressure water at from 270 to 350degC/from 70 to 160 atm even after undergoing neutron irradiation of about 1 x 10 22 n/cm 2 (E>1 MeV) which is a maximum neutron irradiation amount undergone till the final stage of the working life of the LWR-type reactor. In addition, the average thermal expansion coefficient at from room temperature to 400degC ranges from 15x10 -6 - 19x10 -6 /K. (I.N.)

  20. First results of laser welding of neutron irradiated stainless steel

    International Nuclear Information System (INIS)

    Osch, E.V. van; Hulst, D.S. d'; Laan, J.G. van der.

    1994-10-01

    First results of experimental investigations on the laser reweldability of neutron irradiated material are reported. These experiments include the manufacture of 'heterogeneous' joints, which means joining of irradiated stainless steel of type AISI 316L-SPH to 'fresh' unirradiated material. The newly developed laser welding facility in the ECN Hot Cell Laboratory and experimental procedures are described. Visual inspections of welded joints are reported as well as results of electron microscopy and preliminary metallographic examinations. (orig.)

  1. In-reactor precipitation and ferritic transformation in neutron--irradiated stainless steels

    International Nuclear Information System (INIS)

    Porter, D.L.; Wood, E.L.

    1978-01-01

    Ferritic transformation (γ → α) was observed in Type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 400 and 550 0 C. Irradiation-induced segregation of Ni solute to precipitates was found not to influence the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater temperature dependence of precipitation reactions in the process of matrix Ni depletion during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed γ → α transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the γ → α transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present preferential sites for Ni segregation through a defect trapping mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable γ shells around voids in areas of complete transformation

  2. Preliminary microstructural characterization by transmission electron microscopy of 14 MeV neutron irradiated type 316 stainless steel

    International Nuclear Information System (INIS)

    Echer, C.J.

    1977-01-01

    Substantial changes in the mechanical properties of 316 stainless steel were observed after neutron irradiation (phi/sub t/ = 2.3 x 10 21 n/m 2 and E = 14 MeV) at 25 0 C. Comparison of microstructures of the unirradiated and neutron irradiated materials were evaluated using transmission electron microscopy. Evidence of small defect clusters in the irradiated material was found. These findings are consistent with other investigators also evaluating low dose irradiations

  3. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    Science.gov (United States)

    Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

    2012-12-01

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

  4. Self-ion emulation of high dose neutron irradiated microstructure in stainless steels

    Science.gov (United States)

    Jiao, Z.; Michalicka, J.; Was, G. S.

    2018-04-01

    Solution-annealed 304L stainless steel (SS) was irradiated to 130 dpa at 380 °C, and to 15 dpa at 500 °C and 600 °C, and cold-worked 316 SS (CW 316 SS) was irradiated to 130 dpa at 380 °C using 5 MeV Fe++/Ni++ to produce microstructures and radiation-induced segregation (RIS) for comparison with that from neutron irradiation at 320 °C to 46 dpa in the BOR60 reactor. For the 304L SS alloy, self-ion irradiation at 380 °C produced a dislocation loop microstructure that was comparable to that by neutron irradiation. No voids were observed in either the 380 °C self-ion irradiation or the neutron irradiation conditions. Irradiation at 600 °C produced the best match to radiation-induced segregation of Cr and Ni with the neutron irradiation, consistent with the prediction of a large temperature shift by Mansur's invariant relations for RIS. For the CW 316 SS alloy irradiated to 130 dpa at 380 °C, both the irradiated microstructure (dislocation loops, precipitates and voids) and RIS reasonably matched the neutron-irradiated sample. The smaller temperature shift for RIS in CW 316 SS was likely due to the high sink (dislocation) density induced by the cold work. A single self-ion irradiation condition at a dose rate ∼1000× that in reactor does not match both dislocation loops and RIS in solution-annealed 304L SS. However, a single irradiation temperature produced a reasonable match with both the dislocation/precipitate microstructure and RIS in CW 316 SS, indicating that sink density is a critical factor in determining the temperature shift for self-ion irradiations.

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

    International Nuclear Information System (INIS)

    Dalton, J.H.

    1978-01-01

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

  6. In-reactor precipitation and ferritic transformation in neutron-irradiated stainless steels

    International Nuclear Information System (INIS)

    Porter, D.L.; Wood, E.L.

    1979-01-01

    Ferritic transformation (γ→α) was observed in type 304L, 20% cold-worked AISI 316, and solution-annealed AISI 316 stainless steels when subjected to fast neutron irradiation. Each material demonstrated an increasing propensity for transformation with increasing irradiation temperature between 40 and 550 0 C. Irradiation-induced segregation of Ni solute to precipitates was found not to be a controlling factor in the transformation kinetics in 304L. Similar composition data from 316 materials demonstrates a much greater dependence of matrix Ni depletion by precipitation reactions during neutron irradiation. The 316 data establishes a strong link between such depletion and the observed γ→α transformation. Moreover, the lack of correlation between precipitate-related Ni depletion and the γ→α transformation in 304L can be related to the fact that irradiation-induced voids nucleate very quickly in 304L steel during irradiation. These voids present competing sites for Ni segregation through a defect drag mechanism, and hence Ni segregates to voids rather than to precipitates, as evidenced by observed stable γ shells around voids in areas of complete transformation. (Auth.)

  7. Irradiation-induced precipitates in a neutron irradiated 304 stainless steel studied by three-dimensional atom probe

    Energy Technology Data Exchange (ETDEWEB)

    Toyama, T., E-mail: ttoyama@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Nozawa, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Van Renterghem, W. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Matsukawa, Y.; Hatakeyama, M.; Nagai, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Narita-cho 2145-2, Oarai, Ibaraki 311-1313 (Japan); Al Mazouzi, A. [EDF R and D, Avenue des Renardieres Ecuelles, 77818 Moret sur Loing Cedex (France); Van Dyck, S. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium)

    2011-11-15

    Highlights: > Irradiation-induced precipitates in a 304 stainless steel were investigated by three-dimensional atom probe. > The precipitates were found to be {gamma}' precipitates (Ni{sub 3}Si). > Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening. - Abstract: Irradiation-induced precipitates in a 304 stainless steel, neutron-irradiated to a dose of 24 dpa at 300 deg. C in the fuel wrapper plates of a commercial pressurized water reactor, were investigated by laser-assisted three-dimensional atom probe. A high number density of 4 x 10{sup 23} m{sup -3} of Ni-Si rich precipitates was observed, which is one order of magnitude higher than that of Frank loops. The average diameter was {approx}10 nm and the average chemical composition was 40% Ni, 14% Si, 11% Cr and 32% Fe in atomic percent. Over a range of Si concentrations, the ratio of Ni to Si was {approx}3, close to that of {gamma}' precipitate (Ni{sub 3}Si). In some precipitates, Mn enrichment inside the precipitate and P segregation at the interface were observed. Post-irradiation annealing was performed to discuss the contribution of the precipitates to irradiation-hardening.

  8. Effect of phase instabilities on the correlation of nickel ion and neutron irradiation swelling in solution annealed 316 stainless steel

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.; Lee, E.H.; Sklad, P.S.

    1979-01-01

    Annealed 316 stainless steel specimens were neutron irradiated to establish steady-state microstructures and then subjected to further high temperature irradiations with 4 MeV Ni ions. It is shown that void growth under neutron irradiation is simulated in ion irradiations carried out at approx. 180 0 C above reactor temperature. However, the precipitate microstructure developed during neutron irradiation is unstable during subsequent ion irradiation. As a result, the relative swelling rates at various reactor temperatures are not simulated correctly

  9. Microhardness measurement in AISI 321 stainless steel with niobium additions before and after fast neutron irradiation

    International Nuclear Information System (INIS)

    Galli, V.L.; Lucki, G.

    1980-01-01

    Data about influence of neutron irradiation on the microhardness of stainless steel of type AISI 321 with 0.05 and 0.1wt.% Nb additions are presented. The microhardness measurements were made in the range of 300 to 650 0 C, before and after fast neutron irradiation with fluences about 10 17 n/cm 2 . Our results indicate that radiation damage peaks occur around 480 0 C for the stainless steel of type AISI 321 without Nb addition, around 500 0 C for the composition with 0.05 wt.% Nb addition and around 570 0 C for the composition with 0.1 wt.% Nb addition. Microhardness data are in agreement with those obtained by means of electrical resistivity measurements, performed at the same conditions. (Author) [pt

  10. Influence of Silicon on Swelling and Microstructure in Russian Austenitic Stainless Steels Irradiated to High Neutron Doses

    International Nuclear Information System (INIS)

    Porollo, S.I.; Shulepin, S.V.; Konobeev, Y.V.; Garner, F.

    2007-01-01

    Full text of publication follows: For some applications in fusion devices austenitic stainless steels are still considered to be candidates for use as structural components, but high neutron exposures must be endured by the steels. Operational experience of fast reactors in Western Europe, USA and Japan provides evidence of the possible use of austenitic steels up to ∼ 150 dpa. Studies aimed at improvement of existing Russian austenitic steels are being carried out in Russia. For improvement of irradiation resistance of Russian steels it is necessary to understand the basic mechanisms responsible for deterioration of steel properties. This understanding can be achieved by continuing detailed investigations of the microstructure of cladding steels after irradiation to high doses. By investigating the evolution of radiation-induced microstructure in neutron irradiated steels of different chemical composition one can study the effect of chemical variations on steel properties. Silicon is one of the most important chemical elements that strongly influence the behavior of austenitic steel properties under irradiation. In this paper results are presented of investigations of the effect of silicon additions on void swelling and microstructure of base austenitic stainless steel EI-847 (0.06C-16Cr-15Ni- 3Mo-Nb) irradiated as fuel pin cladding of both regular and experimental assemblies in the BOR-60, BN-350 and BN-600 fast reactors to neutron doses up to 49 dpa. The possible mechanisms of silicon's effect on void swelling in austenitic stainless steels are presented and analyzed. (authors)

  11. Effect of titanium on microstructural changes in SUS 316 stainless steels

    International Nuclear Information System (INIS)

    Kawanishi, H.; Yamada, M.; Fukuya, K.; Ishino, S.

    1982-01-01

    The microstructural changes have been examined in order to study the effect of titanium addition to type 316 stainless steels on void swelling. Titanium ions of 400 keV from an accelerator have been implanted at room temperature to solution treated SUS 316 stainless steels which have the original titanium content of 0.02 wt.% to the concentration increase of titanium by 0.01, 0.02 and 0.1 wt.%. Following the preinjection of 20 at.ppm helium at ambient temperature, 400 keV-aluminium ions have been irradiated to the specimen to 40 dpa at 550, 625 and 675 0 C. The TEM observations have revealed that the void number density is drastically increased in the specimen with the content of implanted titanium of more than 0.01 wt.%, whereas the void diameter is remarkably decreased with the titanium content. (orig.)

  12. Effect of fast-neutron irradiation on plastic deformation of Type 304 stainless steel

    International Nuclear Information System (INIS)

    Yamada, H.

    1978-01-01

    Plastic deformation of EBR-II-irradiated Type 304 stainless steel was investigated by a stress-relaxation method. The stress-strain-rate relationships for the irradiated specimens at room temperature are concave upward, which are similar to those for the unirradiated specimens. However, concave downward behavior in the stress-strain-rate relationships were observed at much lower temperatures for the irradiated specimens in contrast to the unirradiated specimens. These results were analyzed succccessfully using Hart's mechanical equation-of-state concept. It was found that the hardness sigma*, which is the minimum stress necessary for the dislocation to overcome obstacles without thermal activation, increases linearly with fast-neutron fluence. This increase in sigma* is consistent with so-called ''irradiation hardening.'' In addition, resistance to dislocation glide, which is quantitatively measured in terms of sigma 0 , was observed to decrease linearly with fast-neutron fluence. The decrease in sigma 0 can be attributed to a decrease of solute drag due to irradiation-induced solute segregation

  13. The corrosion performance of microcrystalline titanium-modified 316 stainless steel

    International Nuclear Information System (INIS)

    Saito, N.; Searson, P.C.; Latanision, R.M.

    1986-01-01

    The corrosion performance of rapidly solidified (RS), consolidated RS and conventionally processed titanium-modified nuclear grade 316 stainless steel was studied. As-solidified RS foils exhibited general corrosion behavior identical to that of the conventionally processed alloy, but inferior pitting resistance, due to the presence of dendritic microsegregation. The consolidated RS alloy exhibited inferior general and pitting corrosion performance due to the detrimental effect of the prior foil boundary formed during the consolidation process. The results of immersion tests in 6% FeC1 3 .6H 2 O solution showed that pit initiation occured primarily at the prior foil boundaries in the consolidated RS alloy. Studies of sensitization were inconclusive due to preferential attack on prior foil boundaries in the consolidated RS specimens which made the determination of the degree of sensitization difficult. (author)

  14. 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.

  15. Irradiation response in titanium modified austenitic stainless steels prepared by rapid solidification processing. Pt. 3

    International Nuclear Information System (INIS)

    Imeson, D.; Tong, C.H.; Parker, C.A.; Vander Sande, J.B.; Grant, N.J.; Harling, O.K.; Massachusetts Inst. of Tech., Cambridge

    1984-01-01

    Titanium carbide precipitation on dislocations during irradiation and recoil-induced particle dissolution are considered. The outline analysis given indicates that complete swelling suppression may occur in favorable conditions due to a counterbalancing of the effective dislocation interstitial bias. The behavior is, however, not stable against a return to normal swelling levels for type 316 steels. A model is presented which may serve as a basis for the interpretation of some aspects of the irradiation response in this system. (orig.)

  16. Effect of prior cold work on creep properties of a titanium modified austenitic stainless steel

    International Nuclear Information System (INIS)

    Vijayanand, V.D.; Parameswaran, P.; Nandagopal, M.; Panneer Selvi, S.; Laha, K.; Mathew, M.D.

    2013-01-01

    Prior cold worked (PCW) titanium-modified 14Cr–15Ni austenitic stainless steel (SS) is used as a core-structural material in fast breeder reactor because of its superior creep strength and resistance to void swelling. In this study, the influence of PCW in the range of 16–24% on creep properties of IFAC-1 SS, a titanium modified 14Cr–15Ni austenitic SS, at 923 K and 973 K has been investigated. It was found that PCW has no appreciable effect on the creep deformation rate of the steel at both the test temperatures; creep rupture life increased with PCW at 923 K and remained rather unaffected at 973 K. The dislocation structure along with precipitation in the PCW steel was found to change appreciably depending on creep testing conditions. A well-defined dislocation substructure was observed on creep testing at 923 K; a well-annealed microstructure with evidences of recrystallization was observed on creep testing at 973 K

  17. Influence of neutron irradiation at 550C on the properties of austenitic stainless steels

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.

    1981-01-01

    Types 316 and 316 + 0.23 wt % Ti stainless steels and 16-8-2 weldment were irradiated in HFIR at 55 0 C to fluences up to 1.35 x 10 26 neutrons/m 2 ( 0 C strength properties, with the weldments the weakest of the materials. The ductility of all materials was reduced by the irradiation, the uniform elongation to only 0.4% in the cold-worked material. Tests at temperatures above the irradiation temperature showed an approach to unirradiated properties as the temperature was increased from 200 to 600 0 C. Helium embrittlement at 700 0 C severely reduced elongation

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

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2014-01-15

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

  19. Precipitation and cavity formation in austenitic stainless steels during irradiation

    International Nuclear Information System (INIS)

    Lee, E.H.; Rowcliffe, A.F.; Mansur, L.K.

    1982-01-01

    Microstructural evolution in austenitic stainless steels subjected to displacement damage at high temperature is strongly influenced by the interaction between helium atoms and second phase particles. Cavity nucleation occurs by the trapping of helium at partially coherent particle-matrix interfaces. The recent precipitate point defect collector theory describes the more rapid growth of precipitate-attached cavities compared to matrix cavities where the precipitate-matrix interface collects point defects to augment the normal point deflect flux to the cavity. Data are presented which support these ideas. It is shown that during nickel ion irradiation of a titanium-modified stainless steel at 675 0 C the rate of injection of helium has a strong effect on the total swelling and also on the nature and distribution of precipitate phases. (orig.)

  20. Stainless Steel to Titanium Bimetallic Transitions

    Energy Technology Data Exchange (ETDEWEB)

    Kaluzny, J. A. [Fermilab; Grimm, C. [Fermilab; Passarelli, D. [Fermilab

    2015-01-01

    In order to use stainless steel piping in an LCLS-II (Linac Coherent Light Source Upgrade) cryomodule, stainless steel to titanium bimetallic transitions are needed to connect the stainless steel piping to the titanium cavity helium vessel. Explosion bonded stainless steel to titanium transition pieces and bimetallic transition material samples have been tested. A sample transition tube was subjected to tests and x-ray examinations between tests. Samples of the bonded joint material were impact and tensile tested at room temperature as well as liquid helium temperature. The joint has been used successfully in horizontal tests of LCLS-II cavity helium vessels and is planned to be used in LCLS-II cryomodules. Results of material sample and transition tube tests will be presented.

  1. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    Energy Technology Data Exchange (ETDEWEB)

    Sato, K., E-mail: ksato@rri.kyoto-u.ac.jp [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Xu, Q.; Yoshiie, T. [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka 590-0494 (Japan); Dai, Y. [Spallation Neutron Source Division, Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Kikuchi, K. [Frontier Research Center for Applied Atomic Sciences, Ibaraki University, Tokai-mura, Naka-gun, Ibaraki 319-1106 (Japan)

    2012-12-15

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<{approx}0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

  2. Positron annihilation lifetime measurements of austenitic stainless and ferritic/martensitic steels irradiated in the SINQ target irradiation program

    International Nuclear Information System (INIS)

    Sato, K.; Xu, Q.; Yoshiie, T.; Dai, Y.; Kikuchi, K.

    2012-01-01

    Titanium-doped austenitic stainless steel (JPCA) and reduced activated ferritic/martensitic steel (F82H) irradiated with high-energy protons and spallation neutrons were investigated by positron annihilation lifetime measurements. Subnanometer-sized (<∼0.8 nm) helium bubbles, which cannot be observed by transmission electron microscopy, were detected by positron annihilation lifetime measurements for the first time. For the F82H steel, the positron annihilation lifetime of the bubbles decreased with increasing irradiation dose and annealing temperature because the bubbles absorb additional He atoms. In the case of JPCA steel, the positron annihilation lifetime increased with increasing annealing temperature above 773 K, in which case the dissociation of complexes of vacancy clusters with He atoms and the growth of He bubbles was detected. He bubble size and density were also discussed.

  3. Influence of temperature histories during reactor startup periods on microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kasahara, Shigeki, E-mail: kasahara.shigeki@jaea.go.jp [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kitsunai, Yuji [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Chimi, Yasuhiro [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Chatani, Kazuhiro; Koshiishi, Masato [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Nishiyama, Yutaka [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2016-11-15

    This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. One of the temperature profiles was that the specimens experienced neutron irradiation in both reactors, under which the irradiation temperature transiently increased to 290 °C from room temperature with increasing reactor power during reactor startup periods. Another was that the specimens were pre-heated to about 150 °C prior to the irradiation to suppress the transient temperature increase. Tensile tests at 290 °C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Difference of the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. Although influence of neutron irradiation involving transient temperature increase to 290 °C from room temperature on the yield strength and the Vickers hardness is buried in the trend curves of existing data, the influence was also found certainly in increment of in yield strength, existence of modest yield drop, and loss of strain hardening capacity and ductility. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, seemed to have important implications regarding the interpretation of not irradiation hardening, but deformation of the austenitic stainless steel.

  4. Strain hardening and plastic instability properties of austenitic stainless steels after proton and neutron irradiation

    International Nuclear Information System (INIS)

    Byun, T.S.; Farrell, K.; Lee, E.H.; Hunn, J.D.; Mansur, L.K.

    2001-01-01

    Strain hardening and plastic instability properties were analyzed for EC316LN, HTUPS316, and AL6XN austenitic stainless steels after combined 800 MeV proton and spallation neutron irradiation to doses up to 10.7 dpa. The steels retained good strain-hardening rates after irradiation, which resulted in significant uniform strains. It was found that the instability stress, the stress at the onset of necking, had little dependence on the irradiation dose. Tensile fracture stress and strain were calculated from the stress-strain curve data and were used to estimate fracture toughness using an existing model. The doses to plastic instability and fracture, the accumulated doses at which the yield stress reaches instability stress or fracture stress, were predicted by extrapolation of the yield stress, instability stress, and fracture stress to higher dose. The EC316LN alloy required the highest doses for plastic instability and fracture. Plastic deformation mechanisms are discussed in relation to the strain-hardening properties of the austenitic stainless steels

  5. The natural aging of austenitic stainless steels irradiated with fast neutrons

    Science.gov (United States)

    Rofman, O. V.; Maksimkin, O. P.; Tsay, K. V.; Koyanbayev, Ye. T.; Short, M. P.

    2018-02-01

    Much of today's research in nuclear materials relies heavily on archived, historical specimens, as neutron irradiation facilities become ever more scarce. These materials are subject to many processes of stress- and irradiation-induced microstructural evolution, including those during and after irradiation. The latter of these, referring to specimens "naturally aged" in ambient laboratory conditions, receives far less attention. The long and slow set of rare defect migration and interaction events during natural aging can significantly change material properties over decadal timescales. This paper presents the results of natural aging carried out over 15 years on austenitic stainless steels from a BN-350 fast breeder reactor, each with its own irradiation, stress state, and natural aging history. Natural aging is shown to significantly reduce hardness in these steels by 10-25% and partially alleviate stress-induced hardening over this timescale, showing that materials evolve back towards equilibrium even at such a low temperature. The results in this study have significant implications to any nuclear materials research program which uses historical specimens from previous irradiations, challenging the commonly held assumption that materials "on the shelf" do not evolve.

  6. Correlation between locally deformed structure and oxide film properties in austenitic stainless steel irradiated with neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro, E-mail: chimi.yasuhiro@jaea.go.jp [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Kitsunai, Yuji [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Kasahara, Shigeki [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Chatani, Kazuhiro; Koshiishi, Masato [Nippon Nuclear Fuel Development, 2163 Narita-cho, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1313 (Japan); Nishiyama, Yutaka [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2016-07-15

    To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%–2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps. - Highlights: • Visible step structures depend on the neutron dose and the applied strain. • Local strain at grain boundaries was accumulated with the neutron dose. • Oxide thickness increases with neutron dose and local strain at grain boundaries. • No penetrative oxidation was observed along grain boundaries or surface steps.

  7. Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1983-01-01

    Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. The study of creep in stainless steel irradiated with fast neutron and alpha particles

    International Nuclear Information System (INIS)

    Correa, D.A.C.

    1985-01-01

    The objective of the present work is to study the creep behavior of the 316 type stainless steel 50% cold worked in different conditions of temperature and applied stress, after neutron radiation and Alfa particles implantation. For this experiment, non-irradiated samples, samples irradiated in the research reactor IEA-R1 with fast neutron (E≥ MeV) up to a fluence of 8.6.10 17 n/cm 2 , and samples implanted with Alfa particles in the cyclotron CV-28 with Helium concentrations of 5 and 26 appm, were creep tested with applied stresses of the 200-300 MPa at temperatures between 650 0 C and 700 0 C. The deformation versus time curves were plotted and it was observed tha the second stage is not well defined, with the creep rate increasing continuously until the occurrence of failure of the material. The study of the effect of increase from 200 MPa to 300 MPa at the same temperature was performed. It can be concluded that this increase produces an approximately 70% reductions in the fracture time of the material, with practically no influence in the total deformation. Samples were tested at different temperatures (650, 675 and 700 0 C) at a same applied stress (200 MPa). It has been observed that a temperature of 50 0 C produces 98,9% of reduction in the fracture time and almost doubles the total deformation. On neutron irradiated samples, creep tests were performed at the same temperature and stress of the non irradiated samples. Comparing the results obtained a tendency of embrittlement due to the neutron irradiation can be observed; no remarkable structure changes were detected due to small fast neutron. Microstructural and metalographic observations were performed before and after each creep test. (author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

    Toyama, T., E-mail: ttoyama@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nozawa, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Van Renterghem, W. [SCK Bullet CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium); Matsukawa, Y.; Hatakeyama, M.; Nagai, Y. [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Al Mazouzi, A. [EDF R and D, Avenue des Renardieres Ecuelles, 77818 Moret sur Loing Cedex (France); Van Dyck, S. [SCK Bullet CEN, Nuclear Materials Science Institute, Boeretang 200, 2400 Mol (Belgium)

    2012-06-15

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

  11. Changes in grain boundary composition induced by neutron irradiation of austenitic stainless steels

    International Nuclear Information System (INIS)

    Asano, K.; Nakata, K.; Fukuya, K.; Kodama, M.

    1992-01-01

    The radiation induced segregation of solutes to the grain boundary in austenitic stainless steels were studied. Type 304 and type 316 steel samples neutron irradiated at 561K up to 9.2x10 25 n/m 2 were obtained and minute compositional profiles across grain boundaries were examined using an analytical scanning transmission electron microscope equipped with a field emission electron gun. Chromium was slightly enriched at grain boundaries at the lowest irradiation dose but decreased with increasing fluence. Higher fluence irradiation resulted in depletion in chromium and molybdenum, and enrichment in nickel, silicon and phosphorus. These changes in grain boundary chemistry were limited within about 5nm of the boundary. Significant depletion of chromium and enrichment of impurities on the grain boundary occurred at fluences roughly coincidental with that of SCC susceptibility change obtained from another project

  12. Study of irradiation effects in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Etienne, A. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Material Department, University of California, Santa Barbara (United States); Pareige, P.; Radiguet, B. [GPM UMR CNRS 6634, Universite et INSA de Rouen (France); Cunningham, N.J.; Odette, G.O. [Material Department, University of California, Santa Barbara (United States); Pokor, C. [EDF RD, departement MMC, site des Renardieres, Moret-sur-Loing (France)

    2011-07-01

    Chemical analyses using Atom Probe Tomography were performed on a bolt made of cold worked 316 austenitic stainless steel, extracted from the internal structures of a pressurized water reactor after seventeen years of reactor service. The irradiation temperature of these samples was 633 K and the irradiation dose was estimated to 12 dpa. These analyses have shown that neutron irradiation has a strong effect on the intragranular distribution of solute atoms. A very high number density (6.10{sup 23} m{sup -3}) of Ni-Si enriched and Cr-Fe depleted clusters was detected after irradiation. In order to bring complementary experimental results and to determine the mechanism of formation of these Ni-Si nano-clusters, Fe{sup 5+} ion irradiations have been performed on a 316 austenitic stainless steel. As after neutron irradiation, the formation of solute enriched features is observed. Linear features and two kinds of clusters, rounded and torus shaped, are present. Considering that solute enriched features are probably formed by radiation induced segregation on point defect sinks, these different shapes are due to the nature of the sinks where segregation occurs. (authors)

  13. Development of advanced austenitic stainless steels resistant to void swelling under irradiation

    International Nuclear Information System (INIS)

    Rouxel, Baptiste

    2016-01-01

    In the framework of studies about Sodium Fast Reactors (SFR) of generation IV, the CEA is developing new austenitic steel grades for the fuel cladding. These steels demonstrate very good mechanical properties but their use is limited because of the void swelling under irradiation. Beyond a high irradiation dose, cavities appear in the alloys and weaken the material. The reference material in France is a 15Cr/15Ni steel, named AIM1, stabilized with titanium. This study try to understand the role played by various chemical elements and microstructural parameters on the formation of the cavities under irradiation, and contribute to the development of a new grade AIM2 more resistant to swelling. In an analytical approach, model materials were elaborated with various chemical compositions and microstructures. Ten grades were cast with chemical variations in Ti, Nb, Ni and P. Four specific microstructures for each alloy highlighted the effect of dislocations, solutes or nano-precipitates on the void swelling. These materials were characterized using TEM and SANS, before irradiation with Fe"2"+ (2 MeV) ions in the order to simulate the damages caused by neutrons. Comparing the irradiated microstructures, it is demonstrated that the solutes have a dominating effect on the formation of cavities. Specifically titanium in solid solution reduces the swelling whereas niobium does not show this effect. Finally, a matrix enriched by 15% to 25% of nickel is still favorable to limit swelling in these advanced austenitic stainless steels. (author) [fr

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

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zhangbo; Lo, Wei-Yang [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States); Chen, Yiren [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Pakarinen, Janne [Belgian Nuclear Research Center (SCK-CEN), Boeretang 200, B-2400 Mol (Belgium); Wu, Yaqiao [Department of Materials Science and Engineering, Boise State University, Boise, ID 83715 (United States); Center for Advanced Energy Studies, Idaho Falls, ID 83401 (United States); Allen, Todd [Engineering Physics Department, University of Wisconsin, Madison, WI 53706 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Yang, Yong, E-mail: yongyang@ufl.edu [Department of Materials Science and Engineering, Nuclear Engineering Program, University of Florida, Gainesville, FL 32611 (United States)

    2015-11-15

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 10{sup 19} n/cm{sup 2}, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10{sup −9} dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    To enable the life extension of Light Water Reactors (LWRs) beyond 60 years, it is critical to gain adequate knowledge for making conclusive predictions to assure the integrity of duplex stainless steel reactor components, e.g. primary pressure boundary and reactor vessel internal. Microstructural changes in the ferrite of thermally aged, neutron irradiated only, and neutron irradiated after being thermally aged cast austenitic stainless steels (CASS) were investigated using atom probe tomography. The thermal aging was performed at 400 °C for 10,000 h and the irradiation was conducted in the Halden reactor at ∼315 °C to 0.08 dpa (5.6 × 10"1"9 n/cm"2, E > 1 MeV). Low dose neutron irradiation at a dose rate of 5 × 10"−"9 dpa/s was found to induce spinodal decomposition in the ferrite of as-cast microstructure, and further to enhance the spinodal decomposition in the thermally aged cast alloys. Regarding the G-phase precipitates, the neutron irradiation dramatically increases the precipitate size, and alters the composition of the precipitates with increased, Mn, Ni, Si and Mo and reduced Fe and Cr contents. The results have shown that low dose neutron irradiation can further accelerate the degradation of ferrite in a duplex stainless steel at the LWR relevant condition.

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

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

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

  17. Evaluation of radiation-induced sensitization using electrochemical potentiokinetic reactivation technique for austenitic stainless steels

    International Nuclear Information System (INIS)

    Inazumi, T.; Bell, G.E.C.; Hishinuma, A.

    1990-01-01

    The electrochemical potentiokinetic reactivation (EPR) test technique was applied to the determination of sensitization in a neutron-irradiated (420 degree C, 10 dpa) titanium-modified austenitic stainless steel. Miniaturized specimens (3 mm diam by 0.25 mm thick) in solution-annealed and 25% cold-worked conditions were tested. The degree of sensitization (DOS) was calculated in terms of the reactivation charge (Pa). Results indicated the occurrence of radiation-induced sensitization when compared to control specimens thermally aged at the irradiation temperature. Post-EPR examination of the specimen surfaces showed etching across the face of each grain as well as at grain boundaries. This indicates that the Pa value normalized by the total grain boundary area, which is an accepted EPR-DOS criterion, cannot be directly used as an indicator of the DOS to determine the susceptibility of this irradiated material to intergranular stress corrosion cracking (IGSCC). Further investigations are necessary to correlate the results in this study to the IGSCC susceptibility of the irradiated stainless steel. 26 refs., 7 figs., 3 tabs

  18. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  19. Multicycle mechanical performance of titanium and stainless steel transpedicular spine implants.

    Science.gov (United States)

    Pienkowski, D; Stephens, G C; Doers, T M; Hamilton, D M

    1998-04-01

    This was a prospective in vitro study comparing titanium alloy and stainless steel alloy in transpedicular spine implants from two different manufactures. To compare the multicycle mechanical performance of these two alloys, used in each of two different implant designs. Transpedicular spine implants primarily have been manufactured from stainless steel, but titanium alloy offers imaging advantages. However, the notch sensitivity of titanium alloy has caused concern regarding how implants made from this material will compare in stiffness and fatigue life with implants made from stainless steel. Twenty-four implants (two alloys, two designs, six implants per group) were mounted in machined polyethylene wafers and repetitively loaded (up to 1 million cycles) from 80 N to 800 N using a 5-Hertz sinusoidal waveform. Load and displacement data were automatically and periodically sampled throughout the entire test. Implant stiffness increased with cycle load number, reached a steady state, then declined just before fatigue failure. Stiffness varied less in titanium transpedicular spine implants than in their stainless counterparts. All stainless steel implant types were stiffer (steady-state value, P titanium alloy counterparts. One titanium implant design failed with fewer (P stainless steel counterpart, whereas a stainless steel implant of another design failed with fewer (P titanium counterpart. Overall, fatigue life, i.e., the total number of load cycles until failure, was related to implant type (P implant material. A transpedicular spine implant's fatigue lifetime depends on both the design and the material and cannot be judged on material alone. Stainless steel implants are stiffer than titanium alloy implants of equal design and size; however, for those designs in which the fatigue life of the titanium alloy version is superior, enlargement of the implant's components can compensate for titanium's lower modulus of elasticity and result in an implant equally stiff

  20. A D-D neutron generator using a titanium drive-in target

    International Nuclear Information System (INIS)

    Kim, I.J.; Jung, N.S.; Jung, H.D.; Hwang, Y.S.; Choi, H.D.

    2008-01-01

    A D-D neutron generator was developed with an intensity of 10 8 n/s. A helicon plasma ion source was used to produce a large current deuteron beam, and neutrons were generated by irradiating the deuteron beam on a titanium drive-in target made of commercial pure titanium. The neutron generator was test-run for several hundred hours, and the performances were investigated. The available range of the deuteron beam current was 0.8-8 mA and the beam could be accelerated up to 97.5 keV. The maximum neutron generation rate in the test-runs was 1.9 x 10 8 n/s, which was achieved by irradiating a 7.6 mA deuteron beam at 94.0 keV on a 0.5 mm-thick target. The operation of the neutron generator was fairly stable, such that the neutron generation rate was not altered by high voltage breakdowns during the test-runs. Neutron generation efficiency was rated as low as 10% when compared to an ideal case of irradiating a 100% monatomic deuteron beam on a perfect TiD 2 target. Factors causing the low efficiency were suggested and discussed

  1. Irradiation response in titanium modified austenitic stainless steels prepared by rapid solidification processing. Pt. 3. A model for the effect of titanium addition

    Energy Technology Data Exchange (ETDEWEB)

    Imeson, D.; Tong, C.H.; Parker, C.A.; Vander Sande, J.B.; Grant, N.J.; Harling, O.K. (Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Materials Science and Engineering; Massachusetts Inst. of Tech., Cambridge (USA). Nuclear Reactor Lab.)

    1984-05-01

    Titanium carbide precipitation on dislocations during irradiation and recoil-induced particle dissolution are considered. The outline analysis given indicates that complete swelling suppression may occur in favorable conditions due to a counterbalancing of the effective dislocation interstitial bias. The behavior is, however, not stable against a return to normal swelling levels for type 316 steels. A model is presented which may serve as a basis for the interpretation of some aspects of the irradiation response in this system.

  2. Microchemical evolution of neutron-irradiated stainless steel

    International Nuclear Information System (INIS)

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

    1980-04-01

    The precipitates that develop in AISI 316 stainless steel during irradiation play a dominant role in determining the dimensional and mechanical property changes of this alloy. This role is expressed primarily in a large change in matrix composition that alters the diffusional properties of the alloy matrix and also appears to alter the rate of acceptance of point defects at dislocations and voids. The major elemental participants in the evolution have been identified as nickel, silicon, and carbon. The exceptional sensitivity of this evolution to many variables accounts for much of the variability of response exhibited by this alloy in nominally similar irradiations

  3. Study of 316 stainless steel swelling due to neutron irradiation

    International Nuclear Information System (INIS)

    Furutani, Gen; Konishi, Takao

    2000-01-01

    Large stresses will be generated in the austenitic stainless steel core internals of pressurized water reactors (PWRs) if excessive swelling occurs after long periods of operation. As a result, deformation or stress corrosion cracking (SCC) could occur in the core internals. However, data on the swelling of irradiated austenitic stainless steel in actual PWRs is limited. In this study, mechanical tests, measurement of produced helium amount and analysis using transmission electron microscopes were carried out on a cold-worked (CW) 316 stainless steel flux thimble tube irradiated up to approximately 35 dpa in a Japanese PWR. The swelling was evaluated to be approximately 0.02%. This level of swelling was much lower than the swelling of the more than several percent that has been observed in fast breeder reactors. (author)

  4. Precipitate evolution in low-nickel austenitic stainless steels during neutron irradiation at very low dose rates

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Garner, F.; Okita, T.

    2007-01-01

    Full text of publication follows: Not all components of a fusion reactor will be subjected to high atomic displacement rates. Some components outside the plasma containment may experience relatively low displacement rates but data generated under long-term irradiation at low dpa rates is hard to obtain. In another study the neutron-induced microstructural evolution in response to long term irradiation at very low dose rates was studied for a Russian low-nickel austenitic stainless steel that is analogous to AISI 304. The irradiated samples were obtained from an out-of-core anti-crush support column for the BN-600 fast reactor with doses ranging from 1.5 to 22 dpa generated at 3x10 -9 to 4x10 -8 dpa/s. The irradiation temperatures were in a very narrow range of 370-375 deg. C. Microstructural observation showed that in addition to voids and dislocations, an unexpectedly high density of small carbide precipitates was formed that are not usually observed at higher dpa rates in this temperature range. These results required us to ask if such unexpected precipitation was anomalous or was a general feature of low-flux, long-term irradiation. It is shown in this paper that a similar behavior was observed in a western stainless steel, namely AISI 304 stainless steel, irradiated at similar temperatures and dpa rates in the EBR-II fast reactor, indicating that irradiation at low dpa rates for many years leads to a different precipitate microstructure and therefore different associated changes in matrix composition than are generated at higher dpa rates. One consequence of this precipitation is a reduced lattice parameter of the alloy matrix, leading to densification that increases in strength with increasing temperature and dose. A. non-destructive method to evaluate these precipitates is under development and is also discussed in this paper. (authors)

  5. Corrosion-resistant titanium nitride coatings formed on stainless steel by ion-beam-assisted deposition

    International Nuclear Information System (INIS)

    Baba, K.; Hatada, R.

    1994-01-01

    Titanium films 70nm thick were deposited on austenitic type 316L stainless steel substrates, and these specimens were irradiated with titanium ions of energy 70kV at a fluence of 1x10 17 ioncm -2 , using a metal vapor vacuum arc (MEVVA) IV metallic ion source at room temperature. After irradiation, titanium nitride (TiN) films were deposited by titanium evaporation and simultaneous irradiation by a nitrogen ion beam, with transport ratios of Ti to N atoms from 0.5 to 10.0 and an ion acceleration voltage of 2kV. The preferred orientation of the TiN films varied from left angle 200 right angle to left angle 111 right angle normal to the surface when the transport ratio was increased. With the help of Auger electron spectroscopy, interfacial mixing was verified. Nitrogen atoms were present in the state of titanium nitride for all transport ratios from 0.5 up to 10.0. However, the chemical bonding state of titanium changed from titanium nitride to the metallic state with increasing transport ratio Ti/N. The corrosion behavior was evaluated in an aqueous solution of sulfuric acid saturated with oxygen, using multisweep cyclic voltammetry measurements. Thin film deposition of pure titanium and titanium implantation prior to TiN deposition have beneficial effects on the suppression of transpassive chromium dissolution. ((orig.))

  6. Calculation of displacement, gas, and transmutation production in stainless steel irradiated with spallation neutrons

    International Nuclear Information System (INIS)

    Wechsler, M.A.; Ramavarapu, R.; Daugherty, E.L.; Palmer, R.C.; Bullen, D.B.; Sommer, W.F.

    1993-01-01

    Calculations using the high-energy transport code LAHET have been made for the production of displacements, helium gas, and transmuted atoms for stainless steel (Fe-18 wt % Cr-10 wt % Ni) irradiated with spallation neutrons at energies of 100 to 1600 MeV. The damage energy cross section increased from about 250 to 350 b keV for increasing neutron energies from 100 to 1600 MeV with a spallation spectrum average of 281 barns-keV. For a displacement threshold energy of 33 eV, the corresponding spectrum-average displacement cross section is 3400 barns. The PKA spectrum was found to be fairly independent of the incident neutron energy, with an average damage energy of 0.25--0.30 MeV. The helium production cross section increased monotonically with increasing neutron energy, with a spectrum average of 0.32 barns. The maximum transmutation yield was observed near manganese (Z = 25), corresponding to a production cross section of about 0.2 barns. Relevance to fusion materials is discussed

  7. 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.)

  8. The irradiation performance of austenitic stainless steel clade PWR fuel rods

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The steady state irradiation performance of austenitic stainless steel clad pressurized water reactor fuel rods is modeled with fuel performance codes of the FRAP series. These codes, originally developed to model the thermal-mechanical behavior of zircaloy clad fuel rods, are modified to model stainless steel clad fuel rods. The irradiation thermal-mechanical behavior of type 348 stainless steel and zircaloy fuel rods is compared. (author) [pt

  9. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  10. Swelling in neutron-irradiated titanium alloys

    International Nuclear Information System (INIS)

    Peterson, D.T.

    1982-04-01

    Immersion density measurements have been performed on a series of titanium alloys irradiated in EBR-II to a fluence of 5 x 10 22 n/cm 2 (E > 0.1 MeV) at 450 and 550 0 C. The materials irradiated were the near-alpha alloys Ti-6242S and Ti-5621S, the alpha-beta alloy Ti-64, and the beta alloy Ti-38644. Swelling was observed in all alloys with the greater swelling being observed at 550 0 C. Microstructural examination revealed the presence of voids in all alloys. Ti-38644 was found to be the most radiation resistant. Ti-6242S and Ti-5621S also displayed good radiation resistance, whereas considerable swelling and precipitation were observed in Ti-64 at 550 0 C

  11. Fracture toughness behavior of irradiated stainless steel in PWR systems

    Energy Technology Data Exchange (ETDEWEB)

    Xu, H.; Fyfitch, S. [AREVA NP Inc., Lynchburg, Pennsylvania (United States); Tang, H.T. [Electric Power Research Inst., Palo Alto, California (United States)

    2007-07-01

    Data from available research programs were collected and evaluated by the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) to determine the relationship between fracture toughness and neutron fluence for conditions representative of pressurized water reactor (PWR) conditions. It is shown that the reduction of fracture toughness with increasing neutron dose in both boiling water reactors (BWRs) and PWRs is consistent with that observed in fast reactors. The lower bound fracture toughness observed for irradiated stainless steels in PWRs is 38 MPa{radical}m (34.6 ksi{radical}in) at neutron exposures greater than 6.7 X 10{sup 21} n/cm{sup 2} (E > 1.0 MeV) or approximately 10 dpa. For such levels of fracture toughness, it is recommended that linear-elastic fracture mechanics (LEFM) analyses be considered for design and operational analyses. The results from this study can be used by the nuclear industry to assess the effects of irradiation on stainless steels in PWR systems. (author)

  12. Fracture toughness of irradiated stainless steel alloys

    International Nuclear Information System (INIS)

    Mills, W.J.

    1986-01-01

    The postirradiation fracture toughness responses of Types 316 and 304 stainless steel (SS) wrought products, cast CF8 SS and Type 308 SS weld deposit were characterized at 427 0 C using J/sub R/-curve techniques. Fast-neutron irradiation of these alloys caused an order of magnitude reduction in J/sub c/ and two orders of magnitude reduction in tearing modulus at neutron exposures above 10 dpa, where radiation-induced losses in toughness appeared to saturate. Saturation J/sub c/ values for the wrought materials ranged from 28 to 31 kJ/m 2 ; the weld exhibited a saturation level of 11 kJ/m 2 . Maximum allowable flaw sizes for highly irradiated stainless steel components stressed to 90% of the unirradiated yield strength are on the order of 3 cm for the wrought material and 1 cm for the weld. Electron fractographic examination revealed that irradiation displacement damage brought about a transition from ductile microvoid coalescence to channel fracture, associated with local separation along planar deformation bands. The lower saturation toughness value for the weld relative to that for the wrought products was attributed to local failure of ferrite particles ahead of the advancing crack which prematurely initiated channel fracture

  13. Characterization of neutron-irradiated HT-UPS steel by high-energy X-ray diffraction microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xuan, E-mail: xuanzhang@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States); Park, Jun-Sang; Almer, Jonathan [Advanced Photon Source, Argonne National Laboratory, Lemont, IL 60439 (United States); Li, Meimei [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL 60439 (United States)

    2016-04-01

    This paper presents the first measurement of neutron-irradiated microstructure using far-field high-energy X-ray diffraction microscopy (FF-HEDM) in a high-temperature ultrafine-precipitate-strengthened (HT-UPS) austenitic stainless steel. Grain center of mass, grain size distribution, crystallographic orientation (texture), diffraction spot broadening and lattice constant distributions of individual grains were obtained for samples in three different conditions: non-irradiated, neutron-irradiated (3dpa/500 °C), and irradiated + annealed (3dpa/500 °C + 600 °C/1 h). It was found that irradiation caused significant increase in grain-level diffraction spot broadening, modified the texture, reduced the grain-averaged lattice constant, but had nearly no effect on the average grain size and grain size distribution, as well as the grain size-dependent lattice constant variations. Post-irradiation annealing largely reversed the irradiation effects on texture and average lattice constant, but inadequately restored the microstrain.

  14. Effect of phosphorus and boron additions on helium bubble microstructure in titanium-modified austenitic stainless steels

    International Nuclear Information System (INIS)

    Jitsukawa, Shiro; Hojou, Kiichi; Hishinuma, Akimichi

    1992-01-01

    Ti-modified austenitic stainless steels (0.06C-0.5Si-15Cr-15Ni-2Mo-0.2Ti) plus P and/or B additions up to 0.06 wt% and 60 wtppm respectively, were irradiated with 10keV He + ions at a dose rate of 1.8 x 10 18 He + /m 2 s at 923K. Irradiation was performed on foil specimens in a transmission electron microscope equipped with an ion-accelerator up to a fluence of 1 x 10 21 He + /m 2 . Small (5 nm >) bubbles were formed in the depth region ranging between 5 and 15 nm from the foil surface of the specimen irradiated up to a fluence of 9 x 10 18 He + /m 2 . These bubbles grew continuously during the following irradiation. They were often observed to grow by coalescence. Bubble growth by coalescence was suppressed by Ti-modification. Both P and B additions enhanced the suppression effect of Ti-modification. (author)

  15. Swelling behaviors in a fuel assembly for the wrapping wire and duct made of modified 316 austenitic stainless steel

    International Nuclear Information System (INIS)

    Yamagata, Ichiro; Akasaka, Naoaki

    2010-01-01

    Swelling behaviors in the wrapping wire and duct made of modified type 316 austenitic stainless steel were investigated in a fuel assembly irradiated in a fast breeder reactor. The temperature dependence of volumetric swelling was measured in the wrapping wire and the duct, and the peak temperatures of swelling were evaluated. The void distribution in the material was measured by microstructure observation with electron microscopy, and it was found that the voids prefentially grew near the surface. This phenomenon seemed to be caused by a surface effect on the neutron-irradiated materials. (author)

  16. Reactor irradiation and helium-3 effects on mechanical properties of alpha-titanium alloys

    International Nuclear Information System (INIS)

    Tebus, V.N.; Alekseev, Eh.F.; Golikov, I.V.

    1990-01-01

    Dependence of α-titanium alloy mechanical properties on test temperature and neutron fluence is investigated. Irradiation is shown to result in material hardening and in their plasticity reduction, but residual plasticity remains rather high. Additional reduction of plasticity results in helium-3 introduced in materials under irradiation. Restoration of properties is observed at test temperature higher 500 deg C. Irradiation by fast neutrons up to high fluences (1.4·10 23 cm -2 ) results in essential alloy softening

  17. 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)

  18. Influence of sodium evnironment on the uniaxial tensile behavior of titanium modified type 316 stainless steel

    International Nuclear Information System (INIS)

    Natesan, K.; Chopra, O.K.; Kassner, T.F.

    1978-01-01

    True stress-true strain tensile data have been obtained for titanium modified type 316 stainless steel in the solution annealed condition and after exposure to a flowing sodium environment at temperature of 700, 650, 600 and 550 0 C. The specimens were exposed to sodium for times between 120 and 5012 h to produce carbon penetration depths in the range 0.05-0.30 mm. The Voce equation was used to describe tensile flow curves for plastic strains above 0.005. The results showed that, when compared with solution annealed specimens, the tensile flow behavior of the sodium exposed specimens is characterized by a higher strain hardening rate, which decreases rapidly as the flow stress increases. The loss in tensile ductility of the material due to carburization in sodium environment was found to be minimal. (Auth.)

  19. Effects of neutron irradiation and fatigue on ductility of stainless steel DIN-1-4948

    International Nuclear Information System (INIS)

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

    1978-01-01

    Test specimens of stainless steel DIN 1.4948, which is similar to AISI Type 304, have been irradiated at 723 and 823 K up to fluences of 1*10$sup 23$ neutrons (n)*m$sup -2$ and 5*10$sup 24$ n*m$sup -2$ (E>0.1 MeV). These are representative conditions for the reactor vessel and inner components of the liquid metal fast breeder reactor SNR-300 after 16 years of operation. High-temperature (723 to 1023 K) tension tests at strain rates ($epsilon$) from 10$sup -7$ to 10 s$sup -1$ 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$sup -7$ and 7*10$sup -6$ 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 to 10 s$sup -1$ (the range of interest for Bethe-Tait accident analyses). A minimum total elongation value of 6 percent is shown at 923 K. The postirradiation fatigue life is reduced by up to about 50 percent 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. 8 refs

  20. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors

    International Nuclear Information System (INIS)

    Pokor, C.

    2003-01-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  1. Behavior under irradiation of super-mirror for neutron guides

    International Nuclear Information System (INIS)

    N'Guy-Marechal, K.

    1997-10-01

    The aim of this work is to study the aging of NiCx/Ti super-mirror multilayers used in neutron guides under thermal neutron irradiation. These multilayers allow an increase of the apparent critical angle of total reflection by creating constructive interferences. Neutrons fluxes are thus increased in neutron guides made with a super-mirror coating. Thin films of one and ten bilayers have been deposited on a silicon and a borosilicate glass substrate. We have then studied the evolution of their optical, structural and mechanical properties after irradiation and annealing. After irradiation, a decrease in neutron reflectivity has been observed, due to the interdiffusion of both materials: this phenomenon was particularly important in the coatings deposited on a glass substrate. X-ray diffraction and X-ray absorption spectroscopy have shown that the structural evolutions of both nickel and titanium do not depend on the substrate. Nickel layers remain face-centered cubic after treatment, whereas the initially hexagonal closed-packed titanium becomes face-centered cubic with a texture in the [111] direction. This phase transformation has been attributed to the formation of a TiH compound containing as much as 50% hydrogen. Despite these structural changes, stress relaxation has occurred after irradiation in our layers. On the contrary, the mean stress that we have determined in previous samples, elaborated in another laboratory, has increased after irradiation. Comparison of both results shows that stress evolution is linked to the deposition conditions. As stress remains almost unchanged after annealing, we may conclude that only irradiation defects, and not heating, lead to stress evolution. Our samples being very similar to real neutron guides, we can extend the results we have obtained in this work to real super-mirrors. (author)

  2. Mechanism of mechanical property enhancement in nitrogen and titanium implanted 321 stainless steel

    International Nuclear Information System (INIS)

    Xu Ming; Li Liuhe; Liu Youming; Cai Xun; Chen Qiulong; Chu, Paul K.

    2006-01-01

    Ion implantation is a well-known method to modify surface mechanical properties. The improvement of the mechanical properties can usually be attributed to the formation of new strengthening phases, solution strengthening, dislocation strengthening, or grain refinement. However, in many cases, the roles of individual factors are not clear. In this study, we implanted nitrogen and titanium into 321 stainless steel samples to investigate the enhancement mechanism of the mechanical properties. Nano-indentation experiments were conducted to measure the hardness under various loadings. The N and Ti implanted 321 stainless steel samples were found to behave differently in the hardness (GPa) versus depth (nm) diagram. The effects of the radiation damage, solution strengthening, and dispersion strengthening phase were analyzed. Characterization of the modified layers was performed using techniques such as Auger electron spectroscopy (AES) and grazing incidence X-ray diffraction (GIXRD). Transmission electron microscopy (TEM) and X-ray diffraction were also applied to reveal the structure of the untreated 321 stainless steel

  3. Fatigue behavior of Type 316 stainless steel following neutron irradiation inducing helium

    International Nuclear Information System (INIS)

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

    1980-01-01

    Since a tokamak reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially in the first wall and blanket. There has been limited work on fatigue in irradiated alloys but none on irradiated materials containing significant amounts of irradiation-induced helium. To provide scoping data and to study the effects of irradiation on fatigue behavior, 20%-cold-worked type 316 stainless steel from the MFE reference heat was studied

  4. Stress transmission through Ti-Ni alloy, titanium and stainless steel in impact compression test.

    Science.gov (United States)

    Yoneyama, T; Doi, H; Kobayashi, E; Hamanaka, H; Tanabe, Y; Bonfield, W

    2000-06-01

    Impact stress transmission of Ti-Ni alloy was evaluated for biomedical stress shielding. Transformation temperatures of the alloy were investigated by means of DSC. An impact compression test was carried out with use of split-Hopkinson pressure-bar technique with cylindrical specimens of Ti-Ni alloy, titanium and stainless steel. As a result, the transmitted pulse through Ti-Ni alloy was considerably depressed as compared with those through titanium and stainless steel. The initial stress reduction was large through Ti-Ni alloy and titanium, but the stress reduction through Ti-Ni alloy was more continuous than titanium. The maximum value in the stress difference between incident and transmitted pulses through Ti-Ni alloy or titanium was higher than that through stainless steel, while the stress reduction in the maximum stress through Ti-Ni alloy was statistically larger than that through titanium or stainless steel. Ti-Ni alloy transmitted less impact stress than titanium or stainless steel, which suggested that the loading stress to adjacent tissues could be decreased with use of Ti-Ni alloy as a component material in an implant system. Copyright 2000 Kluwer Academic Publishers

  5. Emulation of neutron irradiation effects with protons: validation of principle

    International Nuclear Information System (INIS)

    Was, G.S.; Busby, J.T.; Allen, T.; Kenik, E.A.; Jensson, A.; Bruemmer, S.M.; Gan, J.; Edwards, A.D.; Scott, P.M.; Andreson, P.L.

    2002-01-01

    This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 deg. C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 deg. C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 deg. C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a 'W' shaped profile at 1.0 dpa and then into a 'V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary

  6. Effects of thermal aging and neutron irradiation on the mechanical properties of three-wire stainless steel weld overlay cladding

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.

    1997-05-01

    Thermal aging of three-wire series-arc stainless steel weld overlay cladding at 288 degrees C for 1605 h resulted in an appreciable decrease (16%) in the Charpy V-notch (CVN) upper-shelf energy (USE), but the effect on the 41-J transition temperature shift was very small (3 degrees C). The combined effect of aging and neutron irradiation at 288 degrees C to a fluence of 5 x 10 19 neutrons/cm 2 (> 1 MeV) was a 22% reduction in the USE and a 29 degrees C shift in the 41-J transition temperature. The effect of thermal aging on tensile properties was very small. However, the combined effect of irradiation and aging was an increase in the yield strength (6 to 34% at test temperatures from 288 to -125 degrees C) but no apparent change in ultimate tensile strength or total elongation. Neutron irradiation reduced the initiation fracture toughness (J Ic ) much more than did thermal aging alone. Irradiation slightly decreased the tearing modulus, but no reduction was caused by thermal aging alone. Other results from tensile, CVN, and fracture toughness specimens showed that the effects of thermal aging at 288 or 343 degrees C for 20,000 h each were very small and similar to those at 288 degrees C for 1605 h. The effects of long-term thermal exposure time (50,000 h and greater) at 288 degrees C will be investigated as the specimens become available in 1996 and beyond

  7. Surface modification of 17-4PH stainless steel by DC plasma nitriding and titanium nitride film duplex treatment

    International Nuclear Information System (INIS)

    Qi, F.; Leng, Y.X.; Huang, N.; Bai, B.; Zhang, P.Ch.

    2007-01-01

    17-4PH stainless steel was modified by direct current (DC) plasma nitriding and titanium nitride film duplex treatment in this study. The microstructure, wear resistance and corrosion resistance were characterized by X-ray diffraction (XRD), pin-on-disk tribological test and polarization experiment. The results revealed that the DC plasma nitriding pretreatment was in favor of improving properties of titanium nitride film. The corrosion resistance and wear resistance of duplex treatment specimen was more superior to that of only coated titanium nitride film

  8. Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

    International Nuclear Information System (INIS)

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

    1977-01-01

    A review is given of investigations on the rate of phase changes during neutron and electron irradiation in many different fcc alloys showing either precipitation or ordering. The diffusion rate was determined as a function of the irradiation flux, the irradiation temperature and the irradiation dose. It was found that the radiation enhanced diffusion in all the investigated alloys is nearly temperature independent and linearly dependent on the flux. From these results conclusions were drawn concerning the properties of point defects and diffusion mechanisms rate determining during irradiation, which appears to be of a common nature for fcc alloys having a similar structure to those investigated. It has been recognized that the same dependencies which are found for the diffusion rate were also observed for the irradiation creep rate in stainless steels, as reported in literature. On the basis of this observation a combination of measurements is suggested, of radiation enhanced diffusion and radiation enhanced creep in stainless steel alloys. The diffusion tests will be performed at the Euratom Joint Research Centre in Ispra, Italy, and the irradiation creep tests will be carried out in the High Flux Reactor /9/ of the Euratom Joint Research Centre in Petten, The Netherlands. In order to investigate irradiation creep on many samples at a time two special rigs were developed which are distinguished only by the mode of stress applied to the steel specimens. In the first type of rig about 50 samples can be tested uniaxially under tension with various combinations of irradiation temperature and stress. The second type of rig holds up to 70 samples which are tested in bending, again with various combinations of irradiation temperature and stress

  9. Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

    1978-05-01

    Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

  10. Cytotoxicity difference of 316L stainless steel and titanium reconstruction plate

    Directory of Open Access Journals (Sweden)

    Ni Putu Mira Sumarta

    2011-03-01

    Full Text Available Background: Pure titanium is the most biocompatible material today and used as a gold standard for metallic implants. However, stainless steel is still being used as implants because of its strength, ductility, lower price, corrosion resistant and biocompatibility. Purpose: This study was done to revealed the cytotoxicity difference between reconstruction plate made of 316L stainless steel and of commercially pure (CP titanium in baby hamster kidney-21 (BHK-21 fibroblast culture through MTT assay. Methods: Eight samples were prepared from reconstruction plates made of stainless steel type 316L grade 2 (Coen’s reconstruction plate® that had been cut into cylindrical form of 2 mm in diameter and 3 mm long. The other one were made of CP titanium (STEMA Gmbh® of 2 mm in diameter and 2,2 mm long; and had been cleaned with silica paper and ultrasonic cleaner, and sterilized in autoclave at 121° C for 20 minutes.9 Both samples were bathed into microplate well containing 50 μl of fibroblast cells with 2 x 105 density in Rosewell Park Memorial Institute-1640 (RPMI-1640 media, spinned at 30 rpm for 5 minutes. Microplate well was incubated for 24 and 48 hours in 37° C. After 24 hours, each well that will be read at 24 hour were added with 50 μl solution containing 5mg/ml MTT reagent in phosphate buffer saline (PBS solutions, then reincubated for 4 hours in CO2 10% and 37° C. Colorometric assay with MTT was used to evaluate viability of the cells population after 24 hours. Then, each well were added with 50 μl dimethyl sulfoxide (DMSO and reincubated for 5 minutes in 37° C. the wells were read using Elisa reader in 620 nm wave length. Same steps were done for the wells that will be read in 48 hours. Each data were tabulated and analyzed using independent T-test with significance of 5%. Results: This study showed that the percentage of living fibroblast after exposure to 316L stainless steel reconstruction plate was 61.58% after 24 hours and 62

  11. Evolution of stainless steels in nuclear industry

    International Nuclear Information System (INIS)

    Tavassoli, Farhad

    2010-01-01

    Starting with the stainless steels used in the conventional industry, their adoption and successive evolutions in the nuclear industry, from one generation of nuclear reactors to another, is presented. Specific examples for several steels are given, covering fabrication procedures, qualification methods, property databases and design allowable stresses, to show how the ever-increasing demands for better performance and reliability, in particular under neutron irradiation, have been met. Particular attention is paid to the austenitic stainless steels types 304L, 316L, 316L(N), 316L(N)-IG, titanium stabilized grade 321, precipitation strengthened alloy 800, conventional and low activation ferritic/martensitic steels and their oxygen dispersion strengthening (ODS) derivatives. For each material, the evolution of the associated filler metal and welding techniques are also presented. (author)

  12. Effect of periodic temperature variations on the microstructure of neutron-irradiated metals

    DEFF Research Database (Denmark)

    Zinkle, S.J.; Hashimoto, N.; Hoelzer, D.T.

    2002-01-01

    Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom.......-induced microstructural features consisted of dislocation loops, stacking fault tetrahedra and voids in the stainless steel, Ti-rich precipitates in the V alloy, and voids (along with a low density of stacking fault tetrahedra) in copper.......Specimens of pure copper, a high purity austenitic stainless steel, and V–4Cr–4Ti were exposed to eight cycles of either constant temperature or periodic temperature variations during neutron irradiation in the High Flux Isotopes Reactor to a cumulative damage level of 4–5 displacements per atom....... Specimens exposed to periodic temperature variations experienced a low temperature (360 °C) during the initial 10% of accrued dose in each of the eight cycles, and a higher temperature (520 °C) during the remaining 90% of accrued dose in each cycle. The microstructures of the irradiated stainless steel...

  13. Aging under irradiation of super-mirrors used in neutron guides

    International Nuclear Information System (INIS)

    N'Guy-Marechal, K.

    1997-01-01

    The aim of this work is to study the aging of NiC x /Ti super-mirror multilayers used in neutron guides under thermal neutron irradiation. These multilayers allow an increase of the apparent critical angle of total reflection by creating constructive interferences. Neutrons fluxes are thus increased in neutron guides made with a super-mirror coating. Thin films of one and ten bilayers have been deposited on a silicon and a borosilicate glass substrate. We have then studied the evolution of their optical, structural and mechanical properties after irradiation and annealing. After irradiation, a decrease in neutron reflectivity has been observed, due to the interdiffusion of both materials: this phenomenon was particularly important in the coatings deposited on a glass substrate. X-ray diffraction and X-ray absorption spectroscopy have shown that the structural evolutions of both nickel and titanium do not depend on the substrate. Nickel layers remain face-centered cubic after treatment, whereas the initially hexagonal closed-packed titanium becomes face-centered cubic with a texture in the [111] direction. This phase transformation has been attributed to the formation of a TiH compound containing as much as 50 % hydrogen. Despite these structural changes, stress relaxation has occurred after irradiation in our layers. On the contrary, then mean stress that we have determined in previous samples, elaborated in another laboratory, has increased after irradiation. Comparison of both results shows that stress evolution is linked to the deposition conditions. As stress remains almost unchanged after annealing, we may conclude that only irradiation defects, and not heating, lead to stress evolution. Our samples being very similar to real neutron guides, we can extend the results we have obtained in this work to real super-mirrors. (author)

  14. Characteristics and Modification of Non-metallic Inclusions in Titanium-Stabilized AISI 409 Ferritic Stainless Steel

    Science.gov (United States)

    Kruger, Dirk; Garbers-Craig, Andrie

    2017-06-01

    This study describes an investigation into the improvement of castability, final surface quality and formability of titanium-stabilized AISI 409 ferritic stainless steel on an industrial scale. Non-metallic inclusions found in this industrially produced stainless steel were first characterized using SEM-EDS analyses through the INCA-Steel software platform. Inclusions were found to consist of a MgO·Al2O3 spinel core, which acted as heterogeneous nucleation site for titanium solubility products. Plant-scale experiments were conducted to either prevent the formation of spinel, or to modify it by calcium treatment. Modification to spherical dual-phase spinel-liquid matrix inclusions was achieved with calcium addition, which eliminated submerged entry nozzle clogging for this grade. Complete modification to homogeneous liquid calcium aluminates was achieved at high levels of dissolved aluminum. A mechanism was suggested to explain the extent of modification achieved.

  15. 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)

  16. The use of titanium and stainless steel in fracture fixation.

    Science.gov (United States)

    Hayes, J S; Richards, R G

    2010-11-01

    The use of metal in fracture fixation has demonstrated unrivalled success for many years owing to its high stiffness, strength, biological toleration and overall reliable function. The most prominent materials used are electropolished stainless steel and commercially pure titanium, along with the more recent emergence of titanium alloys. Despite the many differences between electropolished stainless steel and titanium, both materials provide a relatively predictable clinical outcome, and offer similar success for fulfilling the main biomechanical and biological requirements of fracture fixation despite distinctive differences in implant properties and biological responses. This article explores these differences by highlighting the limitations and advantages of both materials, and addresses how this translates to clinical success.

  17. Microstructural defect evolution in neutronIrradiated 12Cr18Ni9Ti stainless steel during subsequent isochronous annealing

    Energy Technology Data Exchange (ETDEWEB)

    Tsay, K.V.; Maksimkin, O.P.; Turubarova, L.G.; Rofman, O.V. [Institute of Nuclear Physics NNC RK, Almaty (Kazakhstan); Garner, F.A., E-mail: frank.garner@dslextreme.com [Radiation Effects Consulting, Richland, WA (United States)

    2013-08-15

    Transmission electron microscopy and microhardness measurements were used to examine changes in microstructure and associated strengthening induced in austenitic stainless steel 12Cr18Ni9Ti irradiated to ∼0.001 and ∼5 dpa in the WWR-K reactor before and after being subjected to post-irradiation isochronal annealing. The relatively low values of irradiation temperature and dpa rate (∼80 °C and ∼1.2 × 10{sup −8} dpa/s) experienced by this steel allowed characterization of defect microstructures over a wide range of defect ensembles, all at constant composition, produced first by irradiation and then by annealing at temperatures between 450 and 1050 °C. It was shown that the dispersed barrier hardening model with commonly accepted physical properties successfully predicted the observed hardening. It was also observed that when TiC precipitates form at higher annealing temperatures, the alloy does not change in hardness, reflecting a balance between precipitate-hardening and matrix-softening due to removal of solute-strengthening elements titanium and carbon. Such matrix-softening is not often considered in other studies, especially where the contribution of precipitates to hardening is a second-order effect.

  18. Microstructural defect evolution in neutronIrradiated 12Cr18Ni9Ti stainless steel during subsequent isochronous annealing

    International Nuclear Information System (INIS)

    Tsay, K.V.; Maksimkin, O.P.; Turubarova, L.G.; Rofman, O.V.; Garner, F.A.

    2013-01-01

    Transmission electron microscopy and microhardness measurements were used to examine changes in microstructure and associated strengthening induced in austenitic stainless steel 12Cr18Ni9Ti irradiated to ∼0.001 and ∼5 dpa in the WWR-K reactor before and after being subjected to post-irradiation isochronal annealing. The relatively low values of irradiation temperature and dpa rate (∼80 °C and ∼1.2 × 10 −8 dpa/s) experienced by this steel allowed characterization of defect microstructures over a wide range of defect ensembles, all at constant composition, produced first by irradiation and then by annealing at temperatures between 450 and 1050 °C. It was shown that the dispersed barrier hardening model with commonly accepted physical properties successfully predicted the observed hardening. It was also observed that when TiC precipitates form at higher annealing temperatures, the alloy does not change in hardness, reflecting a balance between precipitate-hardening and matrix-softening due to removal of solute-strengthening elements titanium and carbon. Such matrix-softening is not often considered in other studies, especially where the contribution of precipitates to hardening is a second-order effect

  19. Effectiveness of transfixation and length of instrumentation on titanium and stainless steel transpedicular spine implants.

    Science.gov (United States)

    Korovessis, P; Baikousis, A; Deligianni, D; Mysirlis, Y; Soucacos, P

    2001-04-01

    This study compares the effectiveness of transfixation on the stiffness of two pedicle screw-rod constructs of different manufacture, implant design, and alloy, applied in one-and two-level instability. Four screws composed of either stainless steel or Titanium were assembled in pairs to two polymethylmethacrylate blocks to resemble one-and two-level corpectomy models and the construct underwent nondestructive torsional, extension, and flexion loading. In every loading test, each construct was tested using stainless steel or titanium rods of 4.9-mm diameter in two different lengths (short, 10 cm; long, 15 cm), not augmented or augmented with different transfixation devices or a pair of devices. The authors compared the stiffness of stainless steel and titanium constructs without cross-link with the stiffness of that reinforced with single or double Texas Scottish Rite Hospital (TSRH) cross-link, closed new-type cross-link (closed NTC), or open new-type cross-link (open NTC). The results showed that augmentation or no augmentation of short rods conferred significantly more stiffness than that of long rods of the same material in all three loading modes. The closed NTC provided the greatest increase of torsional, extension, and flexion stiffness, and single TSRH provided the least amount of stiffness. Torsional stiffness of short stainless steel rods augmented or not augmented was significantly greater than that of their titanium counterparts. Torsional stiffness of long titanium rods was always greater than that of their stainless steel counterparts. Extension stiffness of short nonaugmented titanium rods was superior to that of long titanium rods, whereas extension stiffness of nonaugmented short and long stainless steel rods was similar. Nonaugmented short titanium rods showed greater flexion stiffness than that of long titanium rods. Long stainless steel rods displayed significantly greater flexion stiffness than did their titanium counterparts. This

  20. Cytotoxicity difference of 316L stainless steel and titanium reconstruction plate

    OpenAIRE

    Ni Putu Mira Sumarta; Coen Pramono Danudiningrat; Ester Arijani Rachmat; Pratiwi Soesilawati

    2011-01-01

    Background: Pure titanium is the most biocompatible material today and used as a gold standard for metallic implants. However, stainless steel is still being used as implants because of its strength, ductility, lower price, corrosion resistant and biocompatibility. Purpose: This study was done to revealed the cytotoxicity difference between reconstruction plate made of 316L stainless steel and of commercially pure (CP) titanium in baby hamster kidney-21 (BHK-21) fibroblast culture through MTT...

  1. In-Pile Tests for IASCC Growth Behavior of Irradiated 316L Stainless Steel under Simulated BWR Condition in JMTR

    Science.gov (United States)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation test plan to evaluate in-situ effects of neutron/γ-ray irradiation on stress corrosion crack (SCC) growth of irradiated stainless steels using the Japan Materials Testing Reactor (JMTR). SCC growth rate and its dependence on electrochemical corrosion potential (ECP) are different between in-pile test and post irradiation examination (PIE). These differences are not fully understood because of a lack of in-pile data. This paper presents a systematic review on SCC growth data of irradiated stainless steels, an in-pile test plan for crack growth of irradiated SUS316L stainless steel under simulated BWR conditions in the JMTR, and the development of the in-pile test techniques.

  2. The influence of mechanical deformation on the irradiation creep of AISI 316 stainless steel irradiated in the EBR-II and FFTF fast reactors

    International Nuclear Information System (INIS)

    Garner, F.A.; Gilbert, E.R.

    2007-01-01

    Irradiation creep of stainless steels is thought not to be very responsive to material and environmental variables. To test this perception earlier unpublished experiments conducted in the EBR-II reactor on AISI 316 have been analyzed. While swelling is dependent on the cold-work level at 400-480 o C, the post-transient irradiation creep rate, often called the creep compliance B0, is not dependent on cold-work level. If the tube reaches pressures on reactor start-up that generate above-yield stresses in unirradiated steel, then plastic strains occur prior to significant irradiation, but the post-transient strain rate is identical to that of material that did not exceed the yield stress on start-up. It is shown that both stress-free and stress-affected swelling are isotropic and that the Soderberg relationship is maintained. At temperatures above ∼540 o C thermal creep and stored energy begin to assert themselves, with creep rates accelerating with cold-work and becoming non-linear with stress. These results are in agreement with a similar study on titanium-modified 316 steel in FFTF. (author)

  3. Tensile property changes of metals and irradiated to low doses with fission, fusion and spallation neutrons

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

    1992-01-01

    The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructures and mechanical properties of metals. Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36-55 C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90 C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa

  4. Attenuation of Neutron and Gamma Radiation by a Composite Material Based on Modified Titanium Hydride with a Varied Boron Content

    Science.gov (United States)

    Yastrebinskii, R. N.

    2018-04-01

    The investigations on estimating the attenuation of capture gamma radiation by a composite neutron-shielding material based on modified titanium hydride and Portland cement with a varied amount of boron carbide are performed. The results of calculations demonstrate that an introduction of boron into this material enables significantly decreasing the thermal neutron flux density and hence the levels of capture gamma radiation. In particular, after introducing 1- 5 wt.% boron carbide into the material, the thermal neutron flux density on a 10 cm-thick layer is reduced by 11 to 176 factors, and the capture gamma dose rate - from 4 to 9 times, respectively. The difference in the degree of reduction in these functionals is attributed to the presence of capture gamma radiation in the epithermal region of the neutron spectrum.

  5. Defect sink characteristics of specific grain boundary types in 304 stainless steels under high dose neutron environments

    International Nuclear Information System (INIS)

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

    2015-01-01

    Radiation induced segregation (RIS) is a well-studied phenomena which occurs in many structurally relevant nuclear materials including austenitic stainless steels. RIS occurs due to solute atoms preferentially coupling with mobile point defect fluxes that migrate and interact with defect sinks. Here, a 304 stainless steel was neutron irradiated up to 47.1 dpa at 320 °C. Investigations into the RIS response at specific grain boundary types were used to determine the sink characteristics of different boundary types as a function of irradiation dose. A rate theory model built on the foundation of the modified inverse Kirkendall (MIK) model is proposed and benchmarked to the experimental results. This model, termed the GiMIK model, includes alterations in the boundary conditions based on grain boundary structure and expressions for interstitial binding. This investigation, through experiment and modeling, found specific grain boundary structures exhibiting unique defect sink characteristics depending on their local structure. Such interactions were found to be consistent across all doses investigated and to have larger global implications, including precipitation of Ni–Si clusters near different grain boundary types

  6. The "shadow sign": a radiographic differentiation of stainless steel versus titanium spinal instrumentation in spine surgery.

    Science.gov (United States)

    Jones-Quaidoo, Sean M; Novicoff, Wendy; Park, Andrew; Arlet, Vincent

    2011-12-01

    Stainless steel spinal instrumentation has been supplanted in recent years by titanium instrumentation. Knowing whether stainless steel or titanium was used in a previous surgery can guide clinical decision making processes, but frequently the clinician has no way to know what type of metal was used. We describe the radiographic "shadow sign," in which superimposed titanium rods and screws remain radiolucent enough that the contour of the underlying components can be seen on a lateral radiograph, whereas superimposed stainless steel rods and screws are completely radiopaque. This technique was evaluated using a retrospective, randomized, and blinded radiographic comparison of titanium and stainless steel spinal instrumentation. The objective was to determine whether the "shadow sign" can reliably differentiate titanium from stainless steel spinal instrumentation. Lateral radiographs from 16 cases of posterior spinal instrumentation (6 titanium, 6 stainless steel, and 2 replicates of each to assess intraobserver reliability) were randomly selected from a database of cases performed for pediatric scoliosis in a university setting from 2005 to 2009. The cases were randomized then shown to 19 orthopaedic surgery residents, 1 spine fellow, and 2 spine attendings. After the "shadow sign" was described, the surgeons were asked to determine what type of metal each implant was made of. The κ value for both stainless steel and titanium versus the gold standard was 0.83 [standard error (SE) = 0.053], indicating excellent agreement. The κ value for agreement between raters was 0.71 (SE = 0.016) and the κ value for agreement within raters was 0.70 (SE = 0.016), both of which indicated substantial agreement. The "shadow sign" can help a clinician differentiate titanium from stainless steel spinal instrumentation based on radiographic appearance alone. Furthermore, our study reveals that the level of experience in diagnosing spinal lateral radiographs also enhances the use of

  7. Irradiation-induced creep in 316 and 304L stainless steels

    International Nuclear Information System (INIS)

    Walters, L.C.; McVay, G.L.; Hudman, G.D.

    1977-01-01

    Recent results are presented from the in-reactor creep experiments that are being conducted by Argonne National Laboratory. The experiments consist of four subassemblies that contain helium-pressurized as well as unstressed capsules of 316 and 304L stainless steels in several metallurgical conditions. Experiments are being irradiated in row 7 of the EBR-II sodium-cooled fast breeder reactor. Three of the subassemblies are being irradiated at temperatures near 400 0 C, and the fourth subassembly is being irradiated at a temperature of 550 0 C. Creep and swelling strains were determined by profilometer measurements on the full length of the capsules after each irradiation cycle. The accumulated neutron dose on the 304L capsules at 385 0 C was 45 dpa; on the 316 capsules at 400 0 C, 40 dpa; and on the 316 capsules at 550 0 C, 25 dpa. It was found that the in-reactor creep rates were linearly dependent on hoop stress, with the exception being capsules of 316 stainless steel that had been given long-term carbide aging treatment and then irradiated at 550 0 C. Those capsules exhibited much higher creep and swelling rates than their unaged counterparts. For the metallurgical conditions where significant swelling was observed (solution-annealed 304L and aged 316 stainless steels), it was found that the in-reactor creep rates were readily fit to a model that related the creep rates to accumulated swelling. Additionally, it was found that the stress-normalized creep rate for 20%-cold-worked 316 stainless steel at a temperature of 550 0 C was 1.6 times that observed at 400 0 C

  8. Modeling the effects of fast-neutron irradiation on the subsequent mechanical behaviour of type 316 stainless steel

    International Nuclear Information System (INIS)

    Dimelfi, R.J.; Kramer, J.M.

    1980-01-01

    A model describing the subsequent mechanical behaviour of type 316 stainless steel exposed to a fast reactor environment is presented. A constitutive law for the flow stress, based on the Voce equation, is introduced. The strain-rate and temperature-dependence of the flow stress is formulated in this equation; and it is assumed that all contributions to changes in the flow stress, at a given temperature and strain-rate, are manifested in a single hardness parameter. Physically-based models for work hardening, irradition hardening, and irradiation softening are presented. By combining the constitutive law with the relations describing the hardness, a system of equations is presented which describes the mechanical response of the material after specified neutron irradiation and thermal exposure. When the model is used to analyze flow stress data on unirradiated material as well as material irradiated over a wide range of fluence and temperature, agreement between predicted and observed behaviour is shown, supporting the assumption of a single flow law incorporating a single hardness parameter. (orig.)

  9. Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

    Science.gov (United States)

    Latha, S.; Mathew, M. D.; Parameswaran, P.; Nandagopal, M.; Mannan, S. L.

    2011-02-01

    14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

  10. Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Latha, S. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Mathew, M.D., E-mail: mathew@igcar.gov.in [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Parameswaran, P.; Nandagopal, M. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu 603 102 (India); Mannan, S.L. [National Engineering College, Kovilpatti, Tamil Nadu 628 503 (India)

    2011-02-28

    14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

  11. Effect of titanium on the creep deformation behaviour of 14Cr-15Ni-Ti stainless steel

    International Nuclear Information System (INIS)

    Latha, S.; Mathew, M.D.; Parameswaran, P.; Nandagopal, M.; Mannan, S.L.

    2011-01-01

    14Cr-15Ni-Ti modified stainless steel alloyed with additions of phosphorus and silicon is a potential candidate material for the future cores of Prototype Fast Breeder Reactor. In order to optimise the titanium content in this steel, creep tests have been conducted on the heats with different titanium contents of 0.18, 0.23, 0.25 and 0.36 wt.% at 973 K at various stress levels. The stress exponents indicated that the rate controlling deformation mechanism was dislocation creep. A peak in the variation of rupture life with titanium content was observed around 0.23 wt.% titanium and the peak was more pronounced at lower stresses. The variation in creep strength with titanium content was correlated with transmission electron microscopic investigations. The peak in creep strength exhibited by the material with 0.23 wt.% titanium is attributed to the higher volume fraction of fine secondary titanium carbide (TiC) precipitates.

  12. The occurrence of an ordered fcc phase in neutron irradiated M316 stainless steel

    International Nuclear Information System (INIS)

    Cawthorne, C.; Brown, C.

    1977-01-01

    A small precipitate giving a superlattice type diffraction pattern has been observed in M316 type stainless steel irradiated in the Dounreay Fast Reactor. The precipitate was observed in cold worked and solution treated samples which were unstressed and irradiated below 540 0 C, but not in those irradiated above this temperature or in the stressed samples. (B.D.)

  13. Evaluation of non-conformities of hip prostheses made of titanium alloys and stainless steel

    International Nuclear Information System (INIS)

    Bezerra, Ewerton de Oliveira Teotonio; Nascimento, Jose Jeferson da Silva; Luna, Carlos Bruno Barreto; Morais, Crislene Rodrigues da Silva; Campos, Karla Valeria Miranda de

    2017-01-01

    A large number of metallic alloys has satisfactory behavior when used to manufacture implants for hip prostheses. However, they must be in conformity with standards, to ensure their quality for long periods without losing its functionality. Therefore, this paper aims to study the non-conformities in two hip prostheses, one of titanium and other stainless steel according to standards. The implants studied passed by X-ray diffraction (XRD), X-ray fluorescence, tensile test and optical microscopy (OM). Specimens for the tensile test were made according to ASTM E 8M, as well, MO samples passed by metallographic procedure. The results evidenced that some chemical compositions showed in relation to the standards. The XRD analysis showed peaks of austenite and absence of ferrite for the stainless steel, while the titanium alloy presents an alpha phase (HCP) more significant than the beta phase (BCC). The stainless steel alloys and titanium have yield strength and tensile strength that meet the standards. On the other hand, the elastic modulus of the titanium alloy and stainless steel, comes to be ten times greater than the human bone. Therefore, the high modulus of elasticity of the alloys, favors bone resorption problems. The stainless steel microstructure is typical of an austenitic matrix, while the titanium alloy presents α + β microstructure. (author)

  14. Response to annealing and reirradiation of AISI 304L stainless steel following initial high-dose neutron irradiation in EBR-II

    International Nuclear Information System (INIS)

    Porter, D.L.; McVay, G.L.; Walters, L.C.

    1980-01-01

    The object of this study was to measure the stability of irradiation-induced microstructure upon annealing and, by selectively annealing out some of these features and reirradiating the material, it was expected that information could be gained concerning the role of microstructural changes in the void swelling process. Transmission electron microscopic examinations of isochronally annealed (200 to 1050 0 C) AISI 304L stainless steel, which had been irradiated at approximately 415 0 C to a fast (E > 0.1 MeV) neutron fluence of approximately 5.1 x 10 26 n/m 2 , verified that the two-stage hardness recovery with temperatures was related to a low temperature annealing of dislocation structures and a higher temperature annealing of voids and solute redistribution

  15. Comparison of swelling for structural materials on neutron and ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, B.A.

    1986-03-01

    The swelling of V-base alloys, Type 316 stainless steel, Fe-25Ni-15Cr alloys, ferritic steels, Cu, Ni, Nb-1% Zr, and Mo on neutron irradiation is compared with the swelling for these materials on ion irradiation. The results of this comparison show that utilization of the ion-irradiation technique provides for a discriminative assessment of the potential for swelling of candidate materials for fusion reactors.

  16. LET spectra behind high-density titanium and stainless steel hip implants irradiated with a therapeutic proton beam

    Czech Academy of Sciences Publication Activity Database

    Oancea, Cristina; Ambrožová, Iva; Popescu, A. I.; Mytsin, G. V.; Vondráček, V.; Davídková, Marie

    2018-01-01

    Roč. 110, č. 3 (2018), s. 7-13 ISSN 1350-4487 R&D Projects: GA MŠk EF16_013/0001677 Institutional support: RVO:61389005 Keywords : proton therapy * metallic hip implant * titanium * stainless steel * track etched detectors Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 1.442, year: 2016

  17. Cracking behavior of thermally aged and irradiated CF-8 cast austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y., E-mail: Yiren_Chen@anl.gov [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Alexandreanu, B.; Chen, W.-Y.; Natesan, K. [Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States); Li, Z.; Yang, Y. [University of Florida, Gainesville, FL 32611 (United States); Rao, A.S. [US Nuclear Regulatory Commission, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2015-11-15

    To assess the combined effect of thermal aging and neutron irradiation on the cracking behavior of CF-8 cast austenitic stainless steel, crack growth rate (CGR) and fracture toughness J-R curve tests were carried out on compact-tension specimens in high-purity water with low dissolved oxygen. Both unaged and thermally aged specimens were irradiated at ∼320 °C to 0.08 dpa. Thermal aging at 400 °C for 10,000 h apparently had no effect on the corrosion fatigue and stress corrosion cracking behavior in the test environment. The cracking susceptibility of CF-8 was not elevated significantly by neutron irradiation at 0.08 dpa. Transgranular cleavage-like cracking was the main fracture mode during the CGR tests, and a brittle morphology of delta ferrite was often seen on the fracture surfaces at the end of CGR tests. The fracture toughness J-R curve tests showed that both thermal aging and neutron irradiation can induce significant embrittlement. The loss of fracture toughness due to neutron irradiation was more pronounced in the unaged than aged specimens. After neutron irradiation, the fracture toughness values of the unaged and aged specimens were reduced to a similar level. G-phase precipitates were observed in the aged and irradiated specimens with or without prior aging. The similar microstructural changes resulting from thermal aging and irradiation suggest a common microstructural mechanism of inducing embrittlement in CF-8.

  18. Clean forming of stainless steel and titanium products by lubricious oxides

    DEFF Research Database (Denmark)

    Heikkilä, Irma; Wadman, Boel; Thoors, Håkan

    2012-01-01

    to industrial forming processes. Preliminary evaluations show a beneficial influence of two oxides types, on stainless steel and on titanium. More work is needed to test the lubricating effect in other forming operations and to analyse the sustainability aspects for products manufactured with this alternative......Big social benefits can be attained through increased use of stainless steel or titanium in new sheet metal applications. Unfortunately, forming of these materials is often a challenging and costly operation, that can lead to environmental and health problems when solving the technical limitations...

  19. Swelling analysis of austenitic stainless steels by means of ion irradiation and kinetic modeling

    International Nuclear Information System (INIS)

    Kohyama, Akira; Donomae, Takako

    1999-03-01

    The influences of irradiation environment on the swelling behavior of austenitic stainless steel has been studied, to aid understanding the origin of the difference in swelling response of PNC316 stainless steel in fuel-pin environment and in materials irradiation capsules, in terms of irradiation conditions, damage mechanism and material conditions. This work focused on the theoretical investigation of the influence of temperature variation on microstructural development of austenitic stainless steels during irradiation, using a kinetic rate theory model. A modeling and calculation on non-steady irradiation effects were first carried out. A fully dynamic model of point defect evolution and extended defect development, which accounts for cascade damage, was developed and successfully applied to simulate the interstitial loop evolution in low temperature regimes. The influence of cascade interstitial clustering on dislocation loop formation has also been assessed. The establishment of a basis for general assessment of non-steady irradiation effects in austenitic stainless steels was advanced. The developed model was applied to evaluate the influences of temperature variation in formerly carried out CMIR and FFTF/MFA-1 FBR irradiation experiments. The results suggested the gradual approach of microstructural features to equilibrium states in all the temperature variation conditions and no sign of anomalous behavior was noted. On the other hand, there is the influence of temperature variation on microstructural development under the neutron irradiation, like CMIR. So there are some possibilities of the work of mechanism which is not taken care on this model, for example the effect of the precipitate behavior which is sensitive to irradiation temperature. (author)

  20. Structural integrity of stainless steel components exposed to neutron irradiation. Change in failure strength of cracked components due to cold working

    International Nuclear Information System (INIS)

    Kamaya, Masayuki; Hojo, Tomohiro; Mochizuki, Masahito

    2015-01-01

    Load carrying capacity of austenitic stainless steel component is increased due to hardening caused by neutron irradiation if no crack is included in the component. On the other hand, if a crack is initiated in the reactor components, the hardening may decrease the load carrying capacity due to reduction in fracture toughness. In this paper, in order to develop a failure assessment procedure of irradiated cracked components, characteristics of change in failure strength of stainless steels due to cold working were investigated. It was experimentally shown that the proof and tensile strengths were increased by the cold working, whereas the fracture toughness was decreased. The fracture strengths of a cylinder with a circumferential surface crack were analyzed using the obtained material properties. Although the cold working altered the failure mode from plastic collapse to the unsteady ductile crack growth, it did not reduce failure strengths even if 50% cold working was applied. The increase in failure strength was caused not only by increase in flow stress but also by reduction in J-integral value, which was brought by the change in stress-strain curve. It was shown that the failure strength of the hardened stainless steel components could be derived by the two-parameter method, in which the change in material properties could be reasonably considered. (author)

  1. Effect of phosphorus on out-of-pile and in-pile behaviour of stabilized austenitic stainless steels

    International Nuclear Information System (INIS)

    Delalande, C.

    1992-02-01

    This work deals with the improvement of swelling resistance for austenitic stainless steels used as fuel pin cladding in Fast Breeder Reactor. The effect of phosphorus addition and multistabilization by Ti and Nb or Ti, Nb and V are studied on Fe-15Cr-15/25Ni based alloys. First, different ageings are performed to verify the stability of dislocation network, main condition of swelling absence at high irradiation temperature (T>550 deg C, and to study the precipitation, especially the one being able to form during irradiation and to control swelling at lower temperature. Then, 1 MeV electron irradiations are performed to estimate the swelling resistance of these multistabilized steels. Furthermore, neutron radiation induced microstructure of phosphorus modified steels already irradiated in reactor give us fundamental informations to predict and explain the effect of phosphorus and multistabilization on the behaviour of the multistabilized steels. Our results show that niobium plays the same role as titanium on the stabilization ratio in steels, but it is present in more phases. Vanadium seems to have less effect on stability of dislocation network and chemical composition of precipitates. Phosphorus increases the stability of dislocation network of multistabilized steels and FeNbP phosphides are observed at high temperature for phosphorus level above 600 ppm. 1 MeV electron irradiations show that multistabilized steels present good swelling resistance. Phosphorus addition increases the swelling resistance of neutron irradiated steels. (Author). refs., figs., tabs

  2. Is galvanic corrosion between titanium alloy and stainless steel spinal implants a clinical concern?

    Science.gov (United States)

    Serhan, Hassan; Slivka, Michael; Albert, Todd; Kwak, S Daniel

    2004-01-01

    Surgeons are hesitant to mix components made of differing metal classes for fear of galvanic corrosion complications. However, in vitro studies have failed to show a significant potential for galvanic corrosion between titanium and stainless steel, the two primary metallic alloys used for spinal implants. Galvanic corrosion resulting from metal mixing has not been described in the literature for spinal implant systems. To determine whether galvanic potential significantly affects in vitro corrosion of titanium and stainless steel spinal implant components during cyclical compression bending. Bilateral spinal implant constructs consisting of pedicle screws, slotted connectors, 6.35-mm diameter rods and a transverse rod connector assembled in polyethylene test blocks were tested in vitro. Two constructs had stainless steel rods with mixed stainless steel (SS-SS) and titanium (SS-Ti) components, and two constructs had titanium rods with mixed stainless steel (Ti-SS) and titanium (Ti-Ti) components. Each construct was immersed in phosphate-buffered saline (pH 7.4) at 37 C and tested in cyclic compression bending using a sinusoidal load-controlling function with a peak load of 300 N and a frequency of 5 Hz until a level of 5 million cycles was reached. The samples were then removed and analyzed visually for evidence of corrosion. In addition, scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS) were used to evaluate the extent of corrosion at the interconnections. None of the constructs failed during testing. Gross observation of the implant components after disassembly revealed that no corrosion had occurred on the surface of the implants that had not been in contact with another component. The Ti-Ti interfaces showed some minor signs of corrosion only detectable using SEM and EDS. The greatest amount of corrosion occurred at the SS-SS interfaces and was qualitatively less at the SS-Ti and Ti-SS interfaces. The results from this study indicate

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

    International Nuclear Information System (INIS)

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

    1986-04-01

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

  4. Microstructural characterization and model of hardening for the irradiated austenitic stainless steels of the internals of pressurized water reactors; Caracterisation microstructurale et modelisation du durcissement des aciers austenitiques irradies des structures internes des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    Pokor, C

    2003-07-01

    The core internals of Pressurized Water Reactors (PWR) are composed of SA 304 stainless steel plates and CW 316 stainless steel bolts. These internals undergo a neutron flux at a temperature between 280 deg C and 380 deg C which modifies their mechanical properties. These modifications are due to the changes in the microstructure of these materials under irradiation which depend on flux, dose and irradiation temperature. We have studied, by Transmission Electron Microscopy, the microstructure of stainless steels SA 304, CW 316 and CW 316Ti irradiated in a mixed flux reactor (OSIRIS at 330 deg C between 0,8 dpa et 3,4 dpa) and in a fast breeder reactor at 330 deg C (BOR-60) up to doses of 40 dpa. Moreover, samples have been irradiated at 375 deg C in a fast breeder reactor (EBR-II) up to doses of 10 dpa. The microstructure of the irradiated stainless steels consists in faulted Frank dislocation loops in the [111] planes of austenitic, with a Burgers vector of [111]. It is possible to find some voids in the solution annealed samples irradiated at 375 deg C. The evolution of the dislocations loops and voids has been simulated with a 'cluster dynamic' model. The fit of the model parameters has allowed us to have a quantitative description of our experimental results. This description of the microstructure after irradiation was coupled together with a hardening model by Frank loops that has permitted us to make a quantitative description of the hardening of SA 304, CW 316 and CW 316Ti stainless steels after irradiation at a certain dose, flux and temperature. The irradiation doses studied grow up to 90 dpa, dose of the end of life of PWR internals. (author)

  5. A comparative wear study of sputtered ZrN coatings on Si and titanium modified stainless steel substrates

    International Nuclear Information System (INIS)

    Singh, Akash; Kuppusami, P.; Thirumurugesan, R.; Mohandas, E.; Geetha, M.; Kamaraj, V.; Kumar, Niranjan

    2010-01-01

    In the present work wear behaviour of ZrN films grown by a pulsed direct current magnetron sputtering method is reported. The films were grown on silicon (100) and titanium modified stainless steel (alloy-D9) substrates by reactive sputtering in a mixture of argon and nitrogen gases. The structural parameters, preferred orientation and crystallite size as a function of substrate temperatures in the range 300-873 K were studied using X-Ray Diffraction. Deposition parameters have been found to influence the growth rate, crystalline structure and surface roughness, which affect the tribological behaviour of the films. A comparative wear study was performed on these substrates with steel and ceramic balls to evaluate the frictional properties of films. The best tribological performance was found for the sample grown with low flow rates of nitrogen (≤ 2 SCCM) at 873K. The coefficient of friction was found to be lower for the films deposited at higher temperature using steel and ceramic balls. This behaviour was correlated with microstructure and deformation behaviour of coatings. (author)

  6. 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.

  7. 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.

  8. Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

    Energy Technology Data Exchange (ETDEWEB)

    Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi [Musashi Inst. of Tech., Tokyo (Japan); Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao

    1996-04-01

    SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young`s modulus lowers. In addition, its composition elements have the large (n,{alpha}) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

  9. Moessbauer spectroscopy of He irradiated austenitic stainless steel SUS304 at low temperature

    International Nuclear Information System (INIS)

    Horii, Kiyomasa; Ishibashi, Tetsu; Toriyama, Tamotsu; Wakabayashi, Hidehiko; Iijima, Hiroshi; Kawasaki, Katsunori; Hayashi, Nobuyuki; Sakamoto, Isao.

    1996-01-01

    SUS 304 austenitic stainless steel causes the magnetic transition at 60 K, and the Young's modulus lowers. In addition, its composition elements have the large (n,α) reaction cross section to high energy neutrons, and helium is apt to be generated, and this is a factor that lowers the material strength. In the He-irradiated parts in austenitic stainless steel, the precursory state of martensite transformation should exist, and its effect is considered to be observable by carrying out low temperature Moessbauer spectroscopy. As to the preparation of He-irradiation samples, the SUS 304 foils used and the irradiation conditions are described. The measurement of low temperature Moessbauer spectra for the samples without irradiation and with irradiation is reported. In order to determine the magnetic transition point, the thermal scanning measurement was carried out for the samples without or with irradiation. The martensite transformation was measured by X-ray diffraction and transmission type Moessbauer spectroscopy. In order to observe the state of the sample surfaces, the measurement by internal conversion electron Moessbauer spectroscopy was performed. These results and the temperature dependence of the Moessbauer spectra for the irradiated parts are reported. (K.I.)

  10. Design and Fabrication of Titanium Target for Portable Neutron Generator

    International Nuclear Information System (INIS)

    Lee, Cheol Ho; Oh, Byunghoon; Chang, Daesik; Jang, Dohyun; In Sang Yeol; Park, Jaewon; Hong, Kwangpyo

    2014-01-01

    For the neutron generator to produce a neutron flux of the above order, a target that produces fast neutrons in the generator plays an important role, and the target is used and applied to develop the generator due to its simplicity and inexpensive. Making suitable targets for neutron production, especially mono-energy neutrons, has always been of interest. These targets have been used for neutron production reaction studies, calibration of detectors, and neutron therapy. Different studies have been carried out on deuterium and tritium for making solid targets to produce mono-energy neutron from D-D and D-T reactions. A lot of investigations have been carried out on solid target properties such as lifetime, thermal stability, neutron yield, and energy. Vaporized zirconium and titanium layers on a high thermal conductivity substrate (Cu, Mo, Ag) have been used as deuterium and tritium absorbing metals. The density of titanium is smaller than zirconium and the range of charged particles in the titanium targets is more than that in zirconium targets. Thus, titanium targets have more neutron yield than zirconium targets in a low energy beam and titanium is usually used to make a target. The titanium target was designed and simulated to determine the suitable thickness of the target. As a result of the simulation, the target was fabricated to generate fast neutrons by the reaction. The thickness of the target was measured using a profiler. The thickness of the two targets is 2.108 and 2.190 μm. The target will be applied to produce neutrons in a neutron generator

  11. Neutron metrology in the HFR. Steel irradiation. R139-801 (SINAS)

    International Nuclear Information System (INIS)

    Ketema, D.J.

    1999-02-01

    The R139-80 series irradiation experiments is part of the NRG materials test programme to evaluate the irradiation behaviour of several types of austenitic stainless steel. Within this programme five R139-80 specimen holders were irradiated in the HFR Petten to different dose levels. This report presents the final metrology results obtained from activation monitors in a specimen holder, coded as R139-801, containing 12 Compact Tension (CT-10 mm) specimens made from the austenitic stainless steel types 308LSXB/TIG and 304-SXB. The R139-801 assembly was irradiated in channel 1 of a TRIO type facility placed in HFR core-position F8. The aim of this irradiation of specimen holder R139-801 was to reach a minimum target damage level of 7.5 dpa for the specimens at a temperature of 335C. The monitor sets are used to calculate the thermal and fast neutron fluences, displacements per atom and the generated helium content. Additionally detailed information concerning an estimation of the fluence and damage doses received by each specimen and its temperature during irradiation are presented. The main results of the thermal and fast neutron fluence measurements are presented. The results indicate that the obtained damage levels in the steel specimens loaded in this specimen holder vary from 5.8 to 7.9 dpa. The temperatures of the specimens during irradiation varied between 304 and 337C. 14 refs

  12. Comparison of radiographic density and compaction index of root canal obturation using nickel titanium or stainless-steel spreaders

    Directory of Open Access Journals (Sweden)

    M. Adel

    2016-08-01

    Full Text Available Background: Both nickel titanium and stainless-steel spreaders are available. The obvious advantage of nickel titanium spreader over stainless steel spreaders is greater penetration in curved canals. Objective: To compare the radiographic density and compaction index of root canal obturation using nickel-titanium or stainless-steel spreaders in curved canals. Methods: In this experimental study the primary weight of 30 acrylic blocks with 45o degrees of apical curvature were measured by a scale (W1. After canals were prepared by step back master apical up to file #30 all blocks were weighed again (W2 and randomly divided in two groups of 15each. All canals were obturated by Cold lateral compaction technique (with nickel-titanium in one group and stainless-steel finger spreaders in another group. After all blocks were reweighed (W3, compaction index (W3-W2/W1-W2 was calculated. One radiograph was taken for each sample. Apical density of the apical third of each canal was measured by digital transmission densitometer. Data were analyzed statistically using T-test. Findings: Mean compaction index for nickel-titanium group was 7.67±2.38 and for stainless-steel group was 9.14±4.06. There was no significant difference between two groups. Mean radiographic density of obturation was 2.05±0.14 in nickel-titanium group and was 2.07±0.21 in stainless-steel group. There was no significant difference between two groups. Conclusion: It is concluded that nickel-titanium spreaders are not superior than stainless-steel spreaders in obturating curved canal.

  13. The Otto Aufranc Award: Enhanced Biocompatibility of Stainless Steel Implants by Titanium Coating and Microarc Oxidation

    Science.gov (United States)

    Lim, Young Wook; Kwon, Soon Yong; Sun, Doo Hoon

    2010-01-01

    Background Stainless steel is one of the most widely used biomaterials for internal fixation devices, but is not used in cementless arthroplasty implants because a stable oxide layer essential for biocompatibility cannot be formed on the surface. We applied a Ti electron beam coating, to form oxide layer on the stainless steel surface. To form a thicker oxide layer, we used a microarc oxidation process on the surface of Ti coated stainless steel. Modification of the surface using Ti electron beam coating and microarc oxidation could improve the ability of stainless steel implants to osseointegrate. Questions/purposes The ability of cells to adhere to grit-blasted, titanium-coated, microarc-oxidated stainless steel in vitro was compared with that of two different types of surface modifications, machined and titanium-coated, and microarc-oxidated. Methods We performed energy-dispersive x-ray spectroscopy and scanning electron microscopy investigations to assess the chemical composition and structure of the stainless steel surfaces and cell morphology. The biologic responses of an osteoblastlike cell line (SaOS-2) were examined by measuring proliferation (cell proliferation assay), differentiation (alkaline phosphatase activity), and attraction ability (cell migration assay). Results Cell proliferation, alkaline phosphatase activity, migration, and adhesion were increased in the grit-blasted, titanium-coated, microarc-oxidated group compared to the two other groups. Osteoblastlike cells on the grit-blasted, titanium-coated, microarc-oxidated surface were strongly adhered, and proliferated well compared to those on the other surfaces. Conclusions The surface modifications we used (grit blasting, titanium coating, microarc oxidation) enhanced the biocompatibility (proliferation and migration of osteoblastlike cells) of stainless steel. Clinical Relevance This process is not unique to stainless steel; it can be applied to many metals to improve their biocompatibility

  14. Irradiation creep and stress-enhanced swelling of Fe-16Cr-15Ni-Nb austenitic stainless steel in BN-350

    Energy Technology Data Exchange (ETDEWEB)

    Vorobjev, A.N.; Porollo, S.I.; Konobeev, Yu.V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

    1997-04-01

    Irradiation creep and void swelling will be important damage processes for stainless steels when subjected to fusion neutron irradiation at elevated temperatures. The absence of an irradiation device with fusion-relevant neutron spectra requires that data on these processes be collected in surrogate devices such as fast reactors. This paper presents the response of an annealed austenitic steel when exposed to 60 dpa at 480{degrees}C and to 20 dpa at 520{degrees}C. This material was irradiated as thin-walled argon-pressurized tubes in the BN-350 reactor located in Kazakhstan. These tubes were irradiated at hoop stresses ranging from 0 to 200 MPa. After irradiation both destructive and non-destructive examination was conducted.

  15. Crystallization of modified hydroxyapatite on titanium implants

    International Nuclear Information System (INIS)

    Golovanova, O A; Izmailov, R R; Zaits, A V; Ghyngazov, S A

    2016-01-01

    Carbonated-hydroxyapatite (CHA) and Si-hydroxyapatite (Si-HA) precipitation have been synthesized from the model bioliquid solutions (synovial fluid and SBF). It is found that all the samples synthesized from the model solutions are single-phase and represent hydroxyapatite. The crystallization of the modified hydroxyapatite on alloys of different composition, roughness and subjected to different treatment techniques was investigated. Irradiation of the titanium substrates with the deposited biomimetic coating can facilitate further growth of the crystal and regeneration of the surface. (paper)

  16. 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])

  17. Study of stress relief cracking in titanium stabilized austenitic stainless steel

    International Nuclear Information System (INIS)

    Chabaud-Reytier, M.

    1999-01-01

    The heat affected zone (HAZ) of titanium stabilised austenitic stainless steel welds (AISI 321) may exhibit a serious form of intercrystalline cracking during service at high temperature. This type of cracking, called 'stress relief cracking', is known to be due to work hardening but also to ageing: a fine and abundant intragranular Ti(C,N) precipitation appears near the fusion line and modifies the mechanical behaviour of the HAZ. This study aims to better know the accused mechanism and to succeed in estimating the risk of such cracking in welded junctions of 321 stainless steel. To analyse this embrittlement mechanism, and to assess the lifetime of real components, different HAZ are simulated by heat treatments applied to the base material which is submitted to various cold rolling and ageing conditions in order to reproduce the HAZ microstructure. Then, we study the effects of work hardening and ageing on the titanium carbide precipitation, on the mechanical (tensile and creep) behaviour of the resulting material and on its stress relief cracking sensitivity. It is shown that work hardening is the main parameter of the mechanism and that ageing do not favour crack initiation although it leads to titanium carbide precipitation. The role of this precipitation is also discussed. Moreover, a creep damage model is identified by a local approach to fracture. Materials sensitive to stress relief cracking are selected. Then, creep tests are carried out on notched bars in order to quantify the intergranular damage of these different materials; afterwards, these measurements are combined with calculated mechanical fields. Finally, it is shown that the model gives good results to assess crack initiation for a compact tension (CT) specimen during relaxation tests, as well as for a notched tubular specimen tested at 600 deg. C under a steady torque. (author)

  18. Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in BWR Environments

    International Nuclear Information System (INIS)

    Chen, Y.; Chopra, O. K.; Gruber, Eugene E.; Shack, William J.

    2010-01-01

    The internal components of light water reactors are exposed to high-energy neutron irradiation and high-temperature reactor coolant. The exposure to neutron irradiation increases the susceptibility of austenitic stainless steels (SSs) to stress corrosion cracking (SCC) because of the elevated corrosion potential of the reactor coolant and the introduction of new embrittlement mechanisms through radiation damage. Various nonsensitized SSs and nickel alloys have been found to be prone to intergranular cracking after extended neutron exposure. Such cracks have been seen in a number of internal components in boiling water reactors (BWRs). The elevated susceptibility to SCC in irradiated materials, commonly referred to as irradiation-assisted stress corrosion cracking (IASCC), is a complex phenomenon that involves simultaneous actions of irradiation, stress, and corrosion. In recent years, as nuclear power plants have aged and irradiation dose increased, IASCC has become an increasingly important issue. Post-irradiation crack growth rate and fracture toughness tests have been performed to provide data and technical support for the NRC to address various issues related to aging degradation of reactor-core internal structures and components. This report summarizes the results of the last group of tests on compact tension specimens from the Halden-II irradiation. The IASCC susceptibility of austenitic SSs and heat-affected-zone (HAZ) materials sectioned from submerged arc and shielded metal arc welds was evaluated by conducting crack growth rate and fracture toughness tests in a simulated BWR environment. The fracture and cracking behavior of HAZ materials, thermally sensitized SSs and grain-boundary engineered SSs was investigated at several doses (3 dpa). These latest results were combined with previous results from Halden-I and II irradiations to analyze the effects of neutron dose, water chemistry, alloy compositions, and welding and processing conditions on IASCC. The

  19. Review of corrosion phenomena on zirconium alloys, niobium, titanium, inconel, stainless steel, and nickel plate under irradiation

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1975-01-01

    The role of nuclear fluxes in corrosion processes was investigated in ATR, ETR, PRTR, and in Hanford production reactors. Major effort was directed to zirconium alloy corrosion parameter studies. Corrosion and hydriding results are reported as a function of oxygen concentration in the coolant, flux level, alloy composition, surface pretreatment, and metallurgical condition. Localized corrosion and hydriding at sites of bonding to dissimilar metals are described. Corrosion behavior on specimens transferred from oxygenated to low-oxygen coolants in ETR and ATR experiments is compared. Mechanism studies suggest that a depression in the corrosion of the Zr--2.5Nb alloy under irradiation is due to radiation-induced aging. The radiation-induced onset of transition on several alloys is in general a gradual process which nucleates locally, causing areas of oxide prosity which eventually encompass the surface. Examination of Zry-2 process tubes reveals that accelerated corrosion has occurred in low-oxygen coolants. Hydrogen contents are relatively low, but show some localized profiles. Gross hydriding has occurred on process tubes containing aluminum spacers, apparently by a galvanic charging mechanism. Titanium paralleled Zry-2 in corrosion behavior under irradiation. Niobium corrosion was variable, but did not appear to be strongly influenced by radiation. Corrosion rates on Inconel and stainless steel were only slightly higher in-flux than out-of-reactor. Corrosion rates on nickel-plated aluminum appeared to vary substantially with preexposure treatments, but the rates generally were accelerated compared to rates on unirradiated coupons. (59 references, 11 tables, 12 figs.)

  20. Difference in metallic wear distribution released from commercially pure titanium compared with stainless steel plates.

    Science.gov (United States)

    Krischak, G D; Gebhard, F; Mohr, W; Krivan, V; Ignatius, A; Beck, A; Wachter, N J; Reuter, P; Arand, M; Kinzl, L; Claes, L E

    2004-03-01

    Stainless steel and commercially pure titanium are widely used materials in orthopedic implants. However, it is still being controversially discussed whether there are significant differences in tissue reaction and metallic release, which should result in a recommendation for preferred use in clinical practice. A comparative study was performed using 14 stainless steel and 8 commercially pure titanium plates retrieved after a 12-month implantation period. To avoid contamination of the tissue with the elements under investigation, surgical instruments made of zirconium dioxide were used. The tissue samples were analyzed histologically and by inductively coupled plasma atomic emission spectrometry (ICP-AES) for accumulation of the metals Fe, Cr, Mo, Ni, and Ti in the local tissues. Implant corrosion was determined by the use of scanning electron microscopy (SEM). With grades 2 or higher in 9 implants, steel plates revealed a higher extent of corrosion in the SEM compared with titanium, where only one implant showed corrosion grade 2. Metal uptake of all measured ions (Fe, Cr, Mo, Ni) was significantly increased after stainless steel implantation, whereas titanium revealed only high concentrations for Ti. For the two implant materials, a different distribution of the accumulated metals was found by histological examination. Whereas specimens after steel implantation revealed a diffuse siderosis of connective tissue cells, those after titanium exhibited occasionally a focal siderosis due to implantation-associated bleeding. Neither titanium- nor stainless steel-loaded tissues revealed any signs of foreign-body reaction. We conclude from the increased release of toxic, allergic, and potentially carcinogenic ions adjacent to stainless steel that commercially pure Ti should be treated as the preferred material for osteosyntheses if a removal of the implant is not intended. However, neither material provoked a foreign-body reaction in the local tissues, thus cpTi cannot be

  1. Structure and mechanical properties of improved cast stainless steels for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Kenik, E.A.; Busby, J.T. [Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, 37831-6064 (United States); Gussev, M.N., E-mail: gussevmn@ornl.gov [Nuclear Fuel & Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN, 37831-6136 (United States); Maziasz, P.J.; Hoelzer, D.T.; Rowcliffe, A.F.; Vitek, J.M. [Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, 37831-6064 (United States)

    2017-01-15

    Casting of stainless steels is a promising and cost saving way of directly producing large and complex structures, such a shield modules or divertors for the ITER. In the present work, a series of modified high-nitrogen cast stainless steels has been developed and characterized. The steels, based on the cast equivalent of the composition of 316 stainless steel, have increased N (0.14–0.36%) and Mn (2–5.1%) content; copper was added to one of the heats. Mechanical tests were conducted with non-irradiated and 0.7 dpa neutron irradiated specimens. It was established that alloying by nitrogen significantly improves the yield stress of non-irradiated steels and the deformation hardening rate. Manganese tended to decrease yield stress but increased radiation hardening. The role of copper on mechanical properties was negligibly small. Analysis of structure was conducted using SEM-EDS and the nature and compositions of the second phases and inclusions were analyzed in detail. No ferrite formation or significant precipitation were observed in the modified steels. It was shown that the modified steels, compared to reference material (commercial cast 316L steel), had better strength level, exhibit significantly reduced elemental inhomogeneity and only minor second phase formation.

  2. In vitro assessment of photocatalytic titanium oxide surface modified stainless steel orthodontic brackets for antiadherent and antibacterial properties against Lactobacillus acidophilus.

    Science.gov (United States)

    Shah, Alok Girish; Shetty, Pradeep Chandra; Ramachandra, C S; Bhat, N Sham; Laxmikanth, S M

    2011-11-01

    To assess the antiadherent and antibacterial properties of surface modified stainless steel orthodontic brackets with photocatalytic titanium oxide (TiO(2)) against Lactobacillus acidophilus. This study was done on 120 specimens of stainless steel preadjusted edgewise appliance (PEA) orthodontic brackets. The specimens were divided into four test groups. Each group consisted of 30 specimens. Groups containing uncoated brackets acted as a control group for their respective experimental group containing coated brackets. Surface modification of brackets was carried out by the radiofrequency (RF) magnetron sputtering method with photocatalytic TiO(2). Brackets then were subjected to microbiological tests for assessment of the antiadherent and antibacterial properties of photocatalytic TiO(2) coating against L acidophilus. Orthodontic brackets coated with photocatalytic TiO(2) showed an antiadherent effect against L acidophilus compared with uncoated brackets. The bacterial mass that was bound to the TiO(2)-coated brackets was less when compared with the uncoated brackets. Furthermore, TiO(2)-coated brackets had a bactericidal effect on L acidophilus, which causes dental caries. Surface modification of orthodontic brackets with photocatalytic TiO(2) can be used to prevent the accumulation of dental plaque and the development of dental caries during orthodontic treatment.

  3. In-pile IASCC growth tests of irradiated stainless steels in JMTR

    Energy Technology Data Exchange (ETDEWEB)

    Chimi, Yasuhiro; Kasahara, Shigeki; Ise, Hideo; Kawaguchi, Yoshihiko; Nakano, Junichi; Nishiyama, Yutaka [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Shibata, Akira; Ohmi, Masao [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Atomic Energy Agency (JAEA) has an in-pile irradiation-assisted stress corrosion cracking (IASCC) test plan to evaluate in-situ effects of neutron/{gamma}-ray irradiation on crack growth of irradiated stainless steels under high-temperature water conditions for commercial boiling water reactors (BWRs) using the Japan Materials Testing Reactor (JMTR). Crack growth rate and its electrochemical corrosion potential (ECP) dependence are different between in-pile test and post irradiation examination (PIE), but these differences are not fully understood. The objectives of the present study are to understand the difference between in-pile and out-of-pile IASCC growth and to confirm the effectiveness of mitigation due to lowering ECP on in-pile crack growth rates. For in-pile crack growth tests, we have selected a large compact tension specimen such as 0.5T-CT because of validity of SCC growth test at a high stress intensity factor (K-value). For loading a 0.5T-CT specimen up to K - 30 MPa {radical}m, we have adopted a lever type loading unit for in-pile crack growth tests in the JMTR. In this report, an in-pile test plan for crack growth of irradiated SUS316L stainless steels under simulated BWR conditions in the JMTR and current status of development of in-pile crack growth test techniques are presented. (author)

  4. Neutron irradiation effect on thermomechanical properties of shape memory alloys

    International Nuclear Information System (INIS)

    Abramov, V.Ya.; Ionajtis, R.R.; Kotov, V.V.; Loguntsev, E.N.; Ushakov, V.P.

    1996-01-01

    Alloys of Ti-Ni, Ti-Ni-Pd, Fe-Mn-Si, Mn-Cu-Cr, Mn-Cu, Cu-Al-Mn, Cu-Al-Ni systems are investigated after irradiation in IVV-2M reactor at various temperatures with neutron fluence of 10 19 - 10 20 cm -2 . The degradation of shape memory effect in titanium nickelide base alloys is revealed after irradiation. Mn-Cu and Mn-Cu-Cr alloys show the best results. Trends in shape memory alloy behaviour depending on irradiation temperature are found. A consideration is given to the possibility of using these alloys for components of power reactor control and protection systems [ru

  5. Characterization and understanding of ion irradiation effect on the microstructure of austenitic stainless steels

    International Nuclear Information System (INIS)

    Volgin, Alexandre

    2012-01-01

    Austenitic stainless steels are widely used in nuclear industry for internal structures. These structures are located close to the fuel assemblies, inside the pressure vessel. The exposure of these elements to high irradiation doses (the accumulated dose, after 40 years of operation, can reach 80 dpa), at temperature close to 350 C, modifies the macroscopic behavior of the steel: hardening, swelling, creep and corrosion are observed. Moreover, in-service inspections of some of the reactor internal structures have revealed the cracking of some baffle bolts. This cracking has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC). In order to understand this complex phenomenon, a first step is to identify the microstructural changes occurring during irradiation, and to understand the mechanisms at the origin of this evolution. In this framework, a large part of the European project 'PERFORM 60' is dedicated to the study of the irradiation damage in austenitic stainless steels. The objective of this PhD work is to bring comprehensive data on the irradiation effects on microstructure. To reach this goal, two model alloys (FeNiCr and FeNiCrSi) and an industrial austenitic stainless steel (316 steel) are studied using Atom Probe Tomography (APT), Transmission Electron Microscope (TEM) and Positron Annihilation Spectroscopy (PAS). They are irradiated by Ni ions in CSNSM (Orsay) at two temperatures (200 and 450 C) and three doses (0.5, 1 and 5 dpa). TEM observations have shown the appearance of dislocation loops, cavities and staking fault tetrahedra. The dislocation loops in 316 steel were preferentially situated in the vicinity of dislocations, while they were randomly distributed in the FeNiCr alloy. APT study has shown the redistribution of Ni and Si under irradiation in FeNiCrSi model alloy and 316 steel, leading to the appearance of (a) Cottrell clouds along dislocation lines, dislocation loops and other non-identified crystalline defects and (b

  6. Prediction of Irradiation Damage by Artificial Neural Network for Austenitic Stainless Steels

    International Nuclear Information System (INIS)

    Kim, Won Sam; Kim, Dae Whan; Hwang, Seong Sik

    2007-01-01

    The internal structures of pressurized water reactors (PWR) located close to the reactor core are used to support the fuel assemblies, to maintain the alignment between assemblies and the control bars and to canalize the primary water. In general these internal structures consist of baffle plates in solution annealed (SA) 304 stainless steel and baffle bolts in cold worked (CW) 316 stainless steel. These components undergo a large neutron flux at temperatures between 280 and 380 .deg. C. Well-controlled irradiation-assisted stress corrosion cracking (IASCC) data from properly irradiated, and properly characterized, materials are sorely lacking due to the experimental difficulties and financial limitations related to working with highly activated materials. In this work, we tried to apply the artificial neural network (ANN) approach, predicted the susceptibility to an IASCC for an austenitic stainless steel SA 304 and CW 316. G.S. Was and J.-P. Massoud experimental data are used. Because there is fewer experimental data, we need to prediction for radiation damage under the internal structure of PWR. Besides, we compared experimental data with prediction data by the artificial neural network

  7. Complications during removal of stainless steel versus titanium nails used for intramedullary nailing of diaphyseal fractures of the tibia.

    Science.gov (United States)

    Seyhan, Mustafa; Guler, Olcay; Mahirogullari, Mahir; Donmez, Ferdi; Gereli, Arel; Mutlu, Serhat

    2018-02-01

    Intramedullary nailing is the treatment of choice for fractures of the tibial shaft, which might necessitate the nail removal due to complications in the long-term. Although considered as a low-risk procedure, intramedullary nail removal is also associated with certain complications. Here, we compared the most commonly used stainless steel and titanium nails with respect to the complications during removal and clinical outcome for intramedullary nailing of diaphyseal fractures of the tibia. Sixty-two patients (26 females, 36 males) were included in this retrospective study. Of the removed nails, 24 were of stainless steel and 38 of titanium. Preoperative and intraoperative parameters, such as implant discomfort, anterior knee pain, operating time and amount of bleeding, and postoperative outcomes were evaluated for each patient. Titanium nail group had more, but not statistically significant, intraoperative complications than stainless steel group during the removal of nails (p = .4498). Operating time and amount of intraoperative bleeding were significantly higher in titanium group than stainless steel group (p = .0306 and p titanium nails than those of stainless steel nails, whereas there was no difference in terms of postoperative SF-36 and KSS scores. In conclusion, although greater bone contact with titanium increases implant stability, nail removal is more difficult, resulting in more longer surgical operation and more intraoperative bleeding. Therefore, we do not recommend titanium nail removal in asymptomatic patients.

  8. Complications during removal of stainless steel versus titanium nails used for intramedullary nailing of diaphyseal fractures of the tibia

    Directory of Open Access Journals (Sweden)

    Mustafa Seyhan

    2018-02-01

    Full Text Available Objectives: Intramedullary nailing is the treatment of choice for fractures of the tibial shaft, which might necessitate the nail removal due to complications in the long-term. Although considered as a low-risk procedure, intramedullary nail removal is also associated with certain complications. Here, we compared the most commonly used stainless steel and titanium nails with respect to the complications during removal and clinical outcome for intramedullary nailing of diaphyseal fractures of the tibia. Patients and methods: Sixty-two patients (26 females, 36 males were included in this retrospective study. Of the removed nails, 24 were of stainless steel and 38 of titanium. Preoperative and intraoperative parameters, such as implant discomfort, anterior knee pain, operating time and amount of bleeding, and postoperative outcomes were evaluated for each patient. Results: Titanium nail group had more, but not statistically significant, intraoperative complications than stainless steel group during the removal of nails (p = .4498. Operating time and amount of intraoperative bleeding were significantly higher in titanium group than stainless steel group (p = .0306 and p < .001, respectively. Preoperative SF-36 physical component and KSS scores were significantly lower in patients who had removal of titanium nails than those of stainless steel nails, whereas there was no difference in terms of postoperative SF-36 and KSS scores. Conclusion: In conclusion, although greater bone contact with titanium increases implant stability, nail removal is more difficult, resulting in more longer surgical operation and more intraoperative bleeding. Therefore, we do not recommend titanium nail removal in asymptomatic patients. Keywords: Fractures of tibial shaft, Removal of intramedullary nailing, Stainless steel nail, Titanium nail

  9. Comparison of stainless steel and titanium alloy orthodontic miniscrew implants: a mechanical and histologic analysis.

    Science.gov (United States)

    Brown, Ryan N; Sexton, Brent E; Gabriel Chu, Tien-Min; Katona, Thomas R; Stewart, Kelton T; Kyung, Hee-Moon; Liu, Sean Shih-Yao

    2014-04-01

    The detailed mechanical and histologic properties of stainless steel miniscrew implants used for temporary orthodontic anchorage have not been assessed. Thus, the purpose of this study was to compare them with identically sized titanium alloy miniscrew implants. Forty-eight stainless steel and 48 titanium alloy miniscrew implants were inserted into the tibias of 12 rabbits. Insertion torque and primary stability were recorded. One hundred grams of tensile force was applied between half of the implants in each group, resulting in 4 subgroups of 24 specimens each. Fluorochrome labeling was administered at weeks 4 and 5. When the rabbits were euthanized at 6 weeks, stability and removal torque were measured in half (ie, 12 specimens) of each of the 4 subgroups. Microdamage burden and bone-to-implant contact ratio were quantified in the other 12 specimens in each subgroup. Mixed model analysis of variance was used for statistical analysis. All implants were stable at insertion and after 6 weeks. The only significant difference was the higher (9%) insertion torque for stainless steel. No significant differences were found between stainless steel and titanium alloy miniscrew implants in microdamage burden and bone-to-implant contact regardless of loading status. Stainless steel and titanium alloy miniscrew implants provide the same mechanical stability and similar histologic responses, suggesting that both are suitable for immediate orthodontic clinical loads. Copyright © 2014 American Association of Orthodontists. Published by Mosby, Inc. All rights reserved.

  10. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation

    Science.gov (United States)

    Lu, Z.; Faulkner, R. G.; Morgan, T. S.

    2008-12-01

    High Cr ferritic/martensitic steels are leading candidates for structural components of future fusion reactors and new generation fission reactors due to their excellent swelling resistance and thermal properties. A commercial grade 12%CrMoVNb ferritic/martensitic stainless steel in the form of parent plate and off-normal weld materials was fast neutron irradiated up to 33 dpa (1.1 × 10 -6 dpa/s) at 400 °C and 28 dpa (1.7 × 10 -6 dpa/s) at 465 °C, respectively. TEM investigation shows that the fully martensitic weld metal transformed to a duplex austenite/ferrite structure due to high fluence neutron irradiation, the austenite was heavily voided (˜15 vol.%) and the ferrite was relatively void-free; whilst no austenite phases were detected in plate steel. Thermodynamic and phase equilibria software MTDATA has been employed for the first time to investigate neutron irradiation-induced phase transformations. The neutron irradiation effect is introduced by adding additional Gibbs free energy into the system. This additional energy is produced by high energy neutron irradiation and can be estimated from the increased dislocation loop density caused by irradiation. Modelling results show that neutron irradiation reduces the ferrite/austenite transformation temperature, especially for high Ni weld metal. The calculated results exhibit good agreement with experimental observation.

  11. High dose radiation damage in nuclear energy structural materials investigated by heavy ion irradiation simulation

    International Nuclear Information System (INIS)

    Zheng Yongnan; Xu Yongjun; Yuan Daqing

    2014-01-01

    Structural materials in ITER, ADS and fast reactor suffer high dose irradiations of neutrons and/or protons, that leads to severe displacement damage up to lOO dpa per year. Investigation of radiation damage induced by such a high dose irradiation has attracted great attention along with the development of nuclear energy facilities of new generation. However, it is deeply hampered for the lacking of high dose neutron and proton sources. Irradiation simulation of heavy ions produced by accelerators opens up an effective way for laboratory investigation of high dose irradiation induced radiation damage encountered in the ITER, ADS, etc. Radiation damage is caused mainly by atomic displacement in materials. The displacement rate of heavy ions is about lO 3 ∼10 7 orders higher than those of neutrons and protons. High displacement rate of heavy ions significantly reduces the irradiation time. The heavy ion irradiation simulation technique (HIIS) technique has been developed at China Institute of Atomic Energy and a series of the HIIS experiments have been performed to investigate radiation damage in stainless steels, tungsten and tantalum at irradiation temperatures from room temperature to 800 ℃ and in the irradiation dose region up to 100 dpa. The experimental results show that he radiation swelling peak for the modified stainless steel appears in the temperature region around 580 ℃ and the radiation damage is more sensitive to the temperature, the size of the radiation induced vacancy cluster or void increase with the increasing of the irradiation dose, and among the three materials the home-made modified stainless steel has the best radiation resistant property. (authors)

  12. Radiation damage in stainless steel under varying temperature neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Naoaki [Kyushu Univ., Kasuga, Fukuoka (Japan). Research Inst. for Applied Mechanics

    1998-03-01

    Microstructural evolution of model alloys of 316SS was examined by neutron irradiation at JMTR under cyclic temperature varying condition. In the case of Fe-16Cr-17Ni, formation of interstitial loops and voids are strongly suppressed by varying the temperature from 473K to 673K. By adding Ti as miner element (0.25wt%), however, abnormal accumulation of vacancies (void swelling of 11%dpa at 0.1dpa) was observed. Theoretical analysis standing on the rate theory of defect clustering and simulation irradiation experiments with heavy ions indicates that the vacancy-rich condition which appears temporally during and after changing the temperature from low to high brings these results. It was also shown that only 1 dpa pre-irradiation at low temperature changes swelling behavior at high temperature above several 10 dpa. The understanding of non-steady-state defect processes under temperature varying irradiation is very important to estimate the radiation damage under fusion environment where short-term and long-term temperature variation is expected. (author)

  13. Electrochemical behaviour of titanium coated stainless steel by r.f. sputtering in synthetic sweat solutions for electrode applications

    International Nuclear Information System (INIS)

    Fonseca, C.; Vaz, F.; Barbosa, M.A.

    2004-01-01

    The r.f. sputtering technique was used to deposit titanium thin films on stainless steel substrates, aiming at the application of the coated samples as skin contact materials for 'dry' active electrodes. In this work the electrochemical behaviour of the coated samples was investigated in synthetic sweat solutions and their performance was compared with that of uncoated stainless steel and bulk titanium. The characterisation of the samples was carried out by electrochemical techniques and scanning electron microscopy. The coated samples displayed corrosion resistance values in synthetic sweat solutions much higher than stainless steel samples and of the same order of the values measured for bulk titanium in the same conditions

  14. Irradiation-induced precipitation in a SUS316 stainless steel using three-dimensional atom probe

    International Nuclear Information System (INIS)

    Hatakeyama, M.; Yamagata, I.

    2013-01-01

    Precipitation and segregation were investigated in a compositionally modified 316 austenitic stainless steel, neutron-irradiated at 862 K using a three-dimensional atom probe. In the solution-annealed specimen, Mo, Ti, Nb, C and P enrichment were observed in a silicide, with nominal composition Fe 3 Cr 2 Ni 2 Mo 2 Si 2 . In a Ti-rich carbide, nominaling Fe 5 Cr 8 Ni 10 Mo 2 Ti 11 Si 2 C 6 , enrichment of Mo, Si, O, and Nb was observed. Radiation-induced segregation (RIS) at the precipitate–matrix interface was also investigated at an atomic scale. RIS of Ni and P atoms, which are undersized in Fe, was also analyzed around the interface of the Ti-rich carbide and matrix. Results suggest that the carbide–matrix interface is a sink with an interstitial bias. In the cold-worked specimen, complex-precipitates consisting of silicide and carbide were formed

  15. Effect of post-weld heat treatment and neutron irradiation on a dissimilar-metal joint between F82H steel and 316L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Haiying, E-mail: haigirl1983@gmail.com [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); Nagasaka, Takuya [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); National Institute for Fusion Science, Toki (Japan); Kometani, Nobuyuki [Nagoya University, Nagoya (Japan); Muroga, Takeo [SOKENDAI - The Graduated University for Advanced Studies, Toki (Japan); National Institute for Fusion Science, Toki (Japan); Guan, Wenhai; Nogami, Shuhei; Yabuuchi, Kiyohiro; Iwata, Takuya; Hasegawa, Akira [Tohoku University, Sendai (Japan); Yamazaki, Masanori [International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University (Japan); Kano, Sho; Satoh, Yuhki; Abe, Hiroaki [Institute for Materials Research, Tohoku University, Sendai (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho (Japan)

    2015-10-15

    Highlights: • Significant hardening after neutron irradiation at 300 °C for 0.1 dpa was found in the fine-grain HAZ of F82H for the dissimilar-metal joint between F82H and 316L. • The possible hardening mechanism was explained from the viewpoint of carbon behavior. • However, the significant hardening did not degrade the impact property significantly. - Abstract: A dissimilar-metal joint between F82H steel and 316L stainless steel was fabricated by using electron beam welding (EBW). By microstructural analysis and hardness test, the heat-affected zone (HAZ) of F82H was classified into interlayer area, fine-grain area, and coarse-carbide area. Post-weld heat treatment (PWHT) was applied to control the hardness of HAZ. After PWHT at 680 °C for 1 h, neutron irradiation at 300 °C with a dose of 0.1 dpa was carried out for the joint in Belgian Reactor II (BR-II). Compared to the base metals (BMs) and weld metal (WM), significant irradiation hardening up to 450HV was found in the fine-grain HAZ of F82H. However, the impact property of F82H-HAZ specimens, which was machined with the root of the V-notch at HAZ of F82H, was not deteriorated obviously in spite of the significant irradiation hardening.

  16. Crystallography and Morphology of MC Carbides in Niobium-Titanium Modified As-Cast HP Alloys

    Science.gov (United States)

    Buchanan, Karl G.; Kral, Milo V.; Bishop, Catherine M.

    2014-07-01

    The microstructures of two as-cast heats of HP alloy stainless steels modified with niobium and titanium were examined with particular attention paid to the interdendritic niobium-titanium-rich carbides formed during solidification of these alloys. Generally, these precipitates obtain a blocky morphology in the as-cast condition. However, the (NbTi)C precipitates may obtain a nodular morphology. To provide further insight to the origin of the two different morphologies obtained by the (NbTi)C precipitates in the HP-NbTi alloy, the microstructure and crystallography of each have been studied in detail using scanning electron microscopy, transmission electron microscopy, various electron diffraction methods (EBSD, SAD, and CBED), and energy-dispersive X-ray spectroscopy.

  17. Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

    International Nuclear Information System (INIS)

    Gan, J.; Stoller, R.E.; Was, G.S.

    1998-01-01

    A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

  18. Treatment of pediatric femoral shaft fractures by stainless steel and titanium elastic nail system: A randomized comparative trial.

    Science.gov (United States)

    Gyaneshwar, Tank; Nitesh, Rustagi; Sagar, Tomar; Pranav, Kothiyal; Rustagi, Nitesh

    2016-08-01

    Literature suggests that the lower modulus of elasticity of titanium makes it ideal for use in children compared with stainless steel. Better fracture stability was observed in association with titanium nails on torsional and axial compression testing. However, stainless steel nails are stiffer than titanium counterparts, which may provide a rigid construct when fixing paediatric femoral shaft fractures. Complications have been observed more frequently by various researchers when titanium nails are used for fracture fixation in patients with increasing age or weight. The concept of this study was to compare the functional outcome after internal fixation with titanium elastic nail system and stainless steel elastic nail system in paediatric femoral shaft fractures. The study was conducted on 34 patients admitted in the department of orthopaedics, LLRM Medical College & SVBP Hospital, Meerut, India from January 2013 to August 2014. We included patients aged 5-12 years with fracture of the femoral shaft, excluding compound fractures, pathological fractures and other lower limb fractures. Patients were treated by titanium (n=17) or stainless steel (n=17) elastic nail system and followed up for one year. The clinical parameters like range of motion at hip and knee joints, time to full weight bearing on the operated limb and radiological parameters like time to union were compared between two groups. A special note was made of intra- and post-operative complications. Functional outcomes were analysed according to Flynn criteria. Based on the Flynn criteria, 59% of patients had excellent results, 41% had satisfactory results, and no one showed poor results. There was no clinically significant difference between the two groups with respect to time to union and full weight bearing. But the incidence of puncture of the opposite cortex while inserting the nail and trying to advance it through the diaphysis during operation is greatly different. Only one such case was observed

  19. Comparison of fracture behavior for low-swelling ferritic and austenitic alloys irradiated in the Fast Flux Test Facility (FFTF) to 180 DPA

    International Nuclear Information System (INIS)

    Huang, F.H.

    1992-02-01

    Fracture toughness testing was conducted to investigate the radiation embrittlement of high-nickel superalloys, modified austenitic steels and ferritic steels. These materials have been experimentally proven to possess excellent resistance to void swelling after high neutron exposures. In addition to swelling resistance, post-irradiation fracture resistance is another important criterion for reactor material selection. By means of fracture mechanics techniques the fracture behavior of those highly irradiated alloys was characterized in terms of irradiation and test conditions. Precipitation-strengthened alloys failed by channel fracture with very low postirradiation ductility. The fracture toughness of titanium-modified austenitic stainless steel D9 deteriorates with increasing fluence to about 100 displacement per atom (dpa), the fluence level at which brittle fracture appears to occur. Ferritic steels such as HT9 are the most promising candidate materials for fast and fusion reactor applications. The upper-shelf fracture toughness of alloy HT9 remained adequate after irradiation to 180 dpa although its ductile- brittle transition temperature (DBTT) shift by low temperature irradiation rendered the material susceptible to brittle fracture at room temperature. Understanding the fracture characteristics under various irradiation and test conditions helps reduce the potential for brittle fracture by permitting appropriate measure to be taken

  20. Estimates of time-dependent fatigue behavior of Type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

    International Nuclear Information System (INIS)

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

    1978-01-01

    Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irradiated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63*10 26 neutrons (n)/m 2 (E>0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20 percent cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradiations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadings ranging from 2 to 5 MW/m 2 were used. 27 refs

  1. SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 different elements plus a stainless steel

    International Nuclear Information System (INIS)

    Maerker, R.E.

    1976-01-01

    The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables

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

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2015-01-15

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

  3. 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

  4. Effect of triple ion beam irradiation on mechanical properties of high chromium austenitic stainless steel

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Futakawa, Masatoshi; Nanjyo, Yoshiyasu; Kiuchi, Kiyoshi; Anegawa, Takefumi

    2003-01-01

    A high-chromium austenitic stainless steel has been developed for an advanced fuel cladding tube considering waterside corrosion and irradiation embrittlement. The candidate material was irradiated in triple ion (Ni, He, H) beam modes at 573 K up to 50 dpa to simulate irradiation damage by neutron and transmutation product. The change in hardness of the very shallow surface layer of the irradiated specimen was estimated from the slope of load/depth-depth curve which is in direct proportion to the apparent hardness of the specimen. Besides, the Swift's power low constitutive equation (σ=A(ε 0 + ε) n , A: strength coefficient, ε 0 : equivalent strain by cold rolling, n: strain hardening exponent) of the damaged parts was derived from the indentation test combined with an inverse analysis using a finite element method (FEM). For comparison, Type304 stainless steel was investigated as well. Though both Type304SS and candidate material were also hardened by ion irradiation, the increase in apparent hardness of the candidate material was smaller than that of Type304SS. The yield stress and uniform elongation were estimated from the calculated constitutive equation by FEM inverse analysis. The irradiation hardening of the candidate material by irradiation can be expected to be lower than that of Type304SS. (author)

  5. Testing of neutron-irradiated ceramic-to-metal seals

    International Nuclear Information System (INIS)

    Brown, R.D.; Clinard, F.W. Jr.; Lopez, M.R.; Martinez, H.; Romero, T.J.; Cook, J.H.; Barr, H.N.; Hittman, F.

    1990-01-01

    This paper reports on ceramic-to-metal seals prepared by sputtering a titanium metallizing layer onto ceramic disks and then brazing to metal tubes. The ceramics used were alumina, MACOR, spinel, AlON, and a mixture of Al 2 O 3 and Si 3 N 4 . Except for the MACOR, which was brazed to a titanium tube, the ceramics were brazed to niobium tubes. The seals were leak tested and then sent to Los Alamos National Laboratory, where they were irradiated using the spallation neutron source at the Los Alamos Meson Physics Facility. Following irradiation for ∼ 90 days to a fluence of 2.8 x 10 23 n/m 2 , the samples were moved to hot cells and again leak tested. Only the MACOR samples showed any measurable leaks. One set of samples was then pressurized to 6.9 MPa (1000 psi) and subsequently leak tested. No leaks were found. Bursting the seals required hydrostatic pressures of at least 34 MPa (5000 psi). The high seal strength and few leaks indicate that ceramic-to-metal seals can resist radiation-induced degradation

  6. IASCC susceptibility under BWR conditions of welded 304 and 347 stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Castano, M.L. [CIEMAT, Complutense 22, 28040 Madrid (Spain); Schaaf, B. van der [NRG, Petten (Netherlands); Roth, A. [Framatome ANP, Erlangen (Germany); Ohms, C. [JRC-IE, Petten (Netherlands); Gavillet, D. [PSI, Villigen (Switzerland); Dyck, S. van [SCK - CEN, Mol (Belgium)

    2004-07-01

    In-service cracking of Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) internal components has been attributed to Irradiation Assisted Stress Corrosion Cracking (IASCC), a high temperature degradation process that austenitic stainless steels exhibit, when subjected to stress and exposed to relatively high fast neutron flux. Most of the cracking incidents in BWRs were associated to the heat-affected zone (HAZ) of welds. Although the maximum end-of- life dose for this structure is about 3 x 10{sup 20} n/cm{sup 2}, below the threshold fluence of 5 x 10{sup 20} n/cm{sup 2} (equivalent to {approx} 1 dpa) for IASCC in BWR of annealed materials, the influence of neutron irradiation in the weld and HAZ is still an open question. As a consequence of the welding process, residual stresses, microstructural and microchemical modifications are expected. In addition, exposure to neutron irradiation can induce variations in the material's characteristics that can modify the stress corrosion resistance of the welded components. While the IASCC susceptibility of base materials is being widely studied in many international projects, the specific conditions of irradiated weldments are rarely assessed. The INTERWELD project, partially financed by the 5. Framework program of the European Commission, was defined to elucidate neutron radiation induced changes in the HAZ of austenitic stainless steel welds that may promote intergranular cracking. To achieve this goal the evolution of residual stresses, microstructure, micro-chemistry, mechanical properties and the stress corrosion behaviour of irradiated materials are being evaluated. Fabrication of appropriate welds of 304 and 347 stainless steels, representative of core components, was performed. These weld materials were irradiated in the High Flux Reactor (HFR) in Petten to two neutron dose levels, i.e. 0.3 and 1 dpa. Complete characterization of the HAZ of both materials, before and after irradiation is

  7. A comparison of torque expression between stainless steel, titanium molybdenum alloy, and copper nickel titanium wires in metallic self-ligating brackets.

    Science.gov (United States)

    Archambault, Amy; Major, Thomas W; Carey, Jason P; Heo, Giseon; Badawi, Hisham; Major, Paul W

    2010-09-01

    The force moment providing rotation of the tooth around the x-axis (buccal-lingual) is referred to as torque expression in orthodontic literature. Many factors affect torque expression, including the wire material characteristics. This investigation aims to provide an experimental study into and comparison of the torque expression between wire types. With a worm-gear-driven torquing apparatus, wire was torqued while a bracket mounted on a six-axis load cell was engaged. Three 0.019 x 0.0195 inch wire (stainless steel, titanium molybdenum alloy [TMA], copper nickel titanium [CuNiTi]), and three 0.022 inch slot bracket combinations (Damon 3MX, In-Ovation-R, SPEED) were compared. At low twist angles (wires were not statistically significant. At twist angles over 24 degrees, stainless steel wire yielded 1.5 to 2 times the torque expression of TMA and 2.5 to 3 times that of nickel titanium (NiTi). At high angles of torsion (over 40 degrees) with a stiff wire material, loss of linear torque expression sometimes occurred. Stainless steel has the largest torque expression, followed by TMA and then NiTi.

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

    Science.gov (United States)

    Stephenson, Kale J.; Was, Gary S.

    2015-01-01

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

  9. Irradiation-induced precipitation in a SUS316 stainless steel using three-dimensional atom probe

    Energy Technology Data Exchange (ETDEWEB)

    Hatakeyama, M., E-mail: hatake@imr.tohoku.ac.jp [International Research Center for Nuclear Materials Science, IMR/Tohoku University, Narita, Oarai, Ibaraki 311-1313 (Japan); Yamagata, I. [Japan Atom Energy Agency, Narita, Oarai, Ibaraki 311-1393 (Japan)

    2013-11-15

    Precipitation and segregation were investigated in a compositionally modified 316 austenitic stainless steel, neutron-irradiated at 862 K using a three-dimensional atom probe. In the solution-annealed specimen, Mo, Ti, Nb, C and P enrichment were observed in a silicide, with nominal composition Fe{sub 3}Cr{sub 2}Ni{sub 2}Mo{sub 2}Si{sub 2}. In a Ti-rich carbide, nominaling Fe{sub 5}Cr{sub 8}Ni{sub 10}Mo{sub 2}Ti{sub 11}Si{sub 2}C{sub 6}, enrichment of Mo, Si, O, and Nb was observed. Radiation-induced segregation (RIS) at the precipitate–matrix interface was also investigated at an atomic scale. RIS of Ni and P atoms, which are undersized in Fe, was also analyzed around the interface of the Ti-rich carbide and matrix. Results suggest that the carbide–matrix interface is a sink with an interstitial bias. In the cold-worked specimen, complex-precipitates consisting of silicide and carbide were formed.

  10. Irradiation damage of ferritic/martensitic steels: Fusion program data applied to a spallation neutron source

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1997-01-01

    Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs

  11. Influence of the austenitic stainless steel microstructure on the void swelling under ion irradiation

    Directory of Open Access Journals (Sweden)

    Rouxel Baptiste

    2016-01-01

    Full Text Available To understand the role of different metallurgical parameters on the void formation mechanisms, various austenitic stainless steels were elaborated and irradiated with heavy ions. Two alloys, in several metallurgical conditions (15Cr/15Ni–Ti and 15Cr/25Ni–Ti, were irradiated in the JANNUS-Saclay facility at 600 °C with 2 MeV Fe2+ ions up to 150 dpa. Resulting microstructures were observed by Transmission Electron Microscopy (TEM. Different effects on void swelling are highlighted. Only the pre-aged samples, which were consequently solute and especially titanium depleted, show cavities. The nickel-enriched matrix shows more voids with a smaller size. Finally, the presence of nano-precipitates combined with a dense dislocation network decreases strongly the number of cavities.

  12. The precipitation behavior of titanium carbide on the surface of SUS 321 stainless steel

    International Nuclear Information System (INIS)

    Yoshihara, Kazuhiro; Nii, Kazuyoshi

    1982-01-01

    The surface composition of SUS 321 stainless steel at high temperatures was observed in vacuum with Auger electron spectroscopy. The precipitation of titanium carbide was found on the surface of SUS 321. The thickness of precipitated titanium carbide layer increased in proportion to the square root of annealing time and became about 0.05 μm after heated at 1100 K for 432 ks. The precipitated titanium carbide was not replaced by the most surface active element sulfur, and remained stable on the surface. The precipitated layer, however, was not even and had many holes about 1 μm in diameter. The bottom of a hole was SUS 321, on which phosphorus, oxygen and sulfur segregated. As the annealing time was prolonged, these segregants were replaced one by one in the order of the surface activity, and finally the most surface active element, sulfur, remained on the bottom of the hole. Moreover, sulfur diffused over the outside of the hole. The precipitation of titanium carbide on the surface occurred according to the following processes: (1) The titanium and carbon which had been dissolved in the bulk diffused onto the surface of the stainless steel. (2) The titanium carbide which had been precipitated in the bulk dissolved because the concentration of titanum and carbon fell under their solubility limits in the bulk. (3) The titanium and carbon diffused onto the surface which was exposed to vacuum. (4) The titanium and carbon recombined into titanium carbide and precipitated on the surface. The growth rate of the thickness of the precipitated layer was controlled by the diffusion of titanium and carbon in the precipitated titanium carbide. (J.P.N.)

  13. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Abou Taleb, W.M. [Alexandria Univ. (Egypt); Madi, N.K.; Kassem, M.E.; El-Khatib, A.M. [Alexandria Univ. (Egypt). Dept. of Physics

    1996-05-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10{sup 9} n/cm{sup 2}. The optical energy gap E{sub op} exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C{sub p} showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author).

  14. Fast neutron irradiation induced changes in the optical and thermal properties of modified polyvinyl chloride

    International Nuclear Information System (INIS)

    Abou Taleb, W.M.; Madi, N.K.; Kassem, M.E.; El-Khatib, A.M.

    1996-01-01

    The effect of both dopant and neutron radiation on the optical and thermal properties of polyvinyl chloride (PVC) has been studied. The doped samples with Pb and Cd were irradiated with a 14 MeV-neutron fluence in the range 7-28.8 x 10 9 n/cm 2 . The optical energy gap E op exhibits a significant dependence on the type of additive and the neutron irradiation fluence. The specific heat at constant pressure C p showed a nonmonotonical change with radiation fluence. The results of this study show that PVC:Pb behaves as a crystalline structure which is only slightly affected by neutron irradiation, while PVC:Cd is highly affected. (author)

  15. Effects of irradiation on initiation and crack-arrest toughness of two high-copper welds and on stainless steel cladding

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

    1990-01-01

    The objective of the study on the high-copper welds is to determine the effect of neutron irradiation on the shift and shape of the ASME K Ic and K Ia toughness curves. Two submerged-arc welds with copper contents of 0.23 and 0.31 wt % were commercially fabricated in 220-mm-thick plate. Compact specimens fabricated from these welds were irradiated at a nominal temperature of 288 degree C to fluences from 1.5 to 1.9 x 10 19 neutrons/cm 2 (>1 MeV). The fracture toughness test results show that the irradiation-induced shifts at 100 MPa/m were greater than the Charpy 41-J shifts by about 11 and 18 degree C. Mean curve fits indicate mixed results regarding curve shape changes, but curves constructed as lower boundaries to the data do indicate curves of lower slopes. A preliminary evaluation of the crack-arrest results shows that the neutron-irradiation induced crack-arrest toughness temperature shift is about the same as the Charpy V-notch impact temperature shift at the 41-J energy level. The shape of the lower bound curves (for the range of test temperatures covered), compared to those of the ASME K Ia curve did not appear to have been altered by the irradiation. Three-wire stainless steel weld overlay cladding was irradiated at 288 degree C to fluences of 2 and 5 x 10 19 neutrons/cm 2 (>1 MeV). Charpy 41-J temperature shifts of 13 and 28 degree C were observed, respectively. For the lower fluence only, 12.7-mm thick compact specimens showed decreases in both J Ic and the tearing modulus. Comparison of the fracture toughness results with typical plate and a low upper-shelf weld reveals that the irradiated stainless steel cladding possesses low ductile initiation fracture toughness comparable to the low upper-shelf weld. 8 refs., 12 figs., 2 tabs

  16. Application of thermoelectricity to NDE of thermally aged cast duplex stainless steels and neutron irradiated ferritic steels

    International Nuclear Information System (INIS)

    Coste, J.F.; Leborgne, J.M.; Massoud, J.P.; Grisot, O.; Miloudi, S.

    1997-10-01

    The thermoelectric power (TEP) of an alloy depends mainly on its temperature, its chemical composition and its atomic arrangement. The TEP measurement technique is used in order to study and follow two degradation phenomena affecting some components of the primary loop of Pressurized Water Reactors (PWR). The first degradation phenomenon is the thermal aging of cast duplex stainless steel components. The de-mixing of the ferritic Fe-Cr-Ni slid solution is responsible for the decreasing of the mechanical characteristics. Laboratory studies have shown the sensitivity of TEP to the de-mixing phenomenon. TEP increases linearly with the ferrite content and with and Arrhenius-type aging parameter depending on time, temperature and activation energy. TEP is also correlated to mechanic characteristics. The second degradation phenomenon is the aging of ferritic steels due to neutron irradiation at about 290 deg C. In this case, the degradation mechanism is the formation of clusters of solute atoms and/or copper rich precipitates that causes the hardening of the material. As a first approach, a study of binary Fe-Cu alloys irradiated by electrons at 288 deg C has revealed the possibility of following the copper depletion of the ferritic matrix. Moreover, the recovery of the mechanical properties of the alloy by annealing can be monitored. Finally, a correlation between Vickers hardness and TEP has been established. (author)

  17. Effect of neutron irradiation on vanadium alloys

    International Nuclear Information System (INIS)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

  18. Effect of neutron irradiation on vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  19. Application of Thin Layer Activation Technique for Wear and Corrosion Studies in Stainless Steel Using Neutron Sources

    International Nuclear Information System (INIS)

    Mohamed, R.F.I.

    2013-01-01

    In this work elemental analysis for three types of stainless steel samples was performed to compare between their compositions. First the stainless samples were analyzed using Energy Dispersive X-ray (EDX) Spectrometer and Inductively Coupled Plasma Atomic Emission Spectrometry (ICP-AES) as conventional tools for elemental analysis. Second, the samples were subjected to detailed neutron activation analysis (NAA) using Pu-Be neutron source with applying γ-rays spectroscopic measurements for the irradiated samples. The first sample was in the form of thin foils. Eight radioactive isotopes were detected in the measured spectra namely 56 Mn, 59 Fe, 58 Co, 60 Co, 24 Na, 187 W, 99 Mo and 51 Cr which resulted from different neutron reactions with this sample. The other two samples were commercial and the NAA results for one of them show that all of the elements reported in the foil sample are the same except the absence of Mo and the presence of Cr.On the other hand the third sample shows a different composition where only Mn, Fe, and Ni were identified from the measured γ- ray spectra. Stacks of irradiated stainless steal foil and pellets were measured to obtain the activity as a function of thickness using the most intense gamma ray lines of the produced radionuclides. The obtained linear activity-thickness relations for the measured radionuclides were fitted to determine the slope and the maximum thickness which can be measured by this technique. The comparison between these curves showed that the most sensitive radioisotope for detecting slight changes in the thickness is 51 Cr which is formed through the 50 Cr (n,γ) 51 Cr reaction.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. An effective tensile strain hardening exponent was also obtained from the data which shows a relative decrease in ductility of steel with increased irradiation damage. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Alexandreanu, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-11-01

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

  2. Chemical coloring on stainless steel by ultrasonic irradiation.

    Science.gov (United States)

    Cheng, Zuohui; Xue, Yongqiang; Ju, Hongbin

    2018-01-01

    To solve the problems of high temperature and non-uniformity of coloring on stainless steel, a new chemical coloring process, applying ultrasonic irradiation to the traditional chemical coloring process, was developed in this paper. The effects of ultrasonic frequency and power density (sound intensity) on chemical coloring on stainless steel were studied. The uniformity of morphology and colors was observed with the help of polarizing microscope and scanning electron microscopy (SEM), and the surface compositions were characterized by X-ray photoelectric spectroscopy (XPS), meanwhile, the wear resistance and the corrosion resistance were investigated, and the effect mechanism of ultrasonic irradiation on chemical coloring was discussed. These results show that in the process of chemical coloring on stainless steel by ultrasonic irradiation, the film composition is the same as the traditional chemical coloring, and this method can significantly enhance the uniformity, the wear and corrosion resistances of the color film and accelerate the coloring rate which makes the coloring temperature reduced to 40°C. The effects of ultrasonic irradiation on the chemical coloring can be attributed to the coloring rate accelerated and the coloring temperature reduced by thermal-effect, the uniformity of coloring film improved by dispersion-effect, and the wear and corrosion resistances of coloring film enhanced by cavitation-effect. Ultrasonic irradiation not only has an extensive application prospect for chemical coloring on stainless steel but also provides an valuable reference for other chemical coloring. Copyright © 2017 Elsevier B.V. All rights reserved.

  3. Neutron irradiation facility and its characteristics

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji

    1995-01-01

    A neutron irradiation facility utilizing spallation reactions with high energy protons is conceived as one of the facilities in 'Proton Engineering center (PEC)' proposed at JAERI. Characteristics of neutron irradiation field of the facility for material irradiation studies are described in terms of material damage parameters, influence of the pulse irradiation, irradiation environments other than neutronics features, etc., comparing with the other sorts of neutron irradiation facilities. Some perspectives for materials irradiation studies using PEC are presented. (author)

  4. Materials irradiation research in neutron science

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    Materials irradiation researches are planned in Neutron Science Research Program. A materials irradiation facility has been conceived as one of facilities in the concept of Neutron Science Research Center at JAERI. The neutron irradiation field of the facility is characterized by high flux of spallation neutrons with very wide energy range up to several hundred MeV, good accessibility to the irradiation field, good controllability of irradiation conditions, etc. Extensive use of such a materials irradiation facility is expected for fundamental materials irradiation researches and R and D of nuclear energy systems such as accelerator-driven incineration plant for long-lifetime nuclear waste. In this paper, outline concept of the materials irradiation facility, characteristics of the irradiation field, preliminary technical evaluation of target to generate spallation neutrons, and materials researches expected for Neutron Science Research program are described. (author)

  5. Damage evaluation of proton irradiated titanium deuteride thin films to be used as neutron production targets

    Science.gov (United States)

    Suarez Anzorena, Manuel; Bertolo, Alma A.; Gagetti, Leonardo; Gaviola, Pedro A.; del Grosso, Mariela F.; Kreiner, Andrés J.

    2018-06-01

    Titanium deuteride thin films have been manufactured under different conditions specified by deuterium gas pressure, substrate temperature and time. The films were characterized by different techniques to evaluate the deuterium content and the homogeneity of such films. Samples with different concentrations of deuterium, including non deuterated samples, were irradiated with a 150 keV proton beam. Both deposits, pristine and irradiated, were characterized by optical profilometry and scanning electron microscopy.

  6. Radiological risks from irradiation of cargo contents with EURITRACK neutron inspection systems

    International Nuclear Information System (INIS)

    Giroletti, E.; Bonomi, G.; Donzella, A.; Viesti, G.; Zenoni, A.

    2012-01-01

    The radiological risk for the population related to the neutron irradiation of cargo containers with a tagged neutron inspection system has been studied. Two possible effects on the public health have been assessed: the modification of the nutritional and organoleptic properties of the irradiated materials, in particular foodstuff, and the neutron activation of consumer products (i.e. food and pharmaceuticals). The result of this study is that irradiation of food and foodstuff, pharmaceutical and medical devices in container cargoes would neither modify the properties of the irradiated material nor produce effective doses of concern for public health. Furthermore, the dose received by possible stowaways present inside the container during the inspection is less than the annual effective dose limit defined by European Legislation for the public. - Highlights: ► Neutron irradiation of cargo containers implies a radiological risk. ► The risk is about the modification of food properties and the products activation. ► Assessment is made about the EURITRACK neutron irradiation system. ► Results show that the EURITRACK scanning is not dangerous for the population.

  7. Irradiated accelerated corrosion of stainless steel

    International Nuclear Information System (INIS)

    Raiman, S.S.; Wang, P.; Was, G.S.

    2015-01-01

    Type 316L stainless steel was exposed to a simulated PWR environment with in-situ proton irradiation to investigate the effect of simultaneous irradiation and corrosion. To enable these experiments, a dedicated beamline was constructed to transport a 3.2 MeV proton beam from a tandem accelerator, through the sample that also acts as the window between the beamline vacuum and a corrosion cell designed to flow primary water at 320 C. degrees and 13.1 MPa. Experiments were conducted on 316L stainless steel samples which were irradiated for 24 hours in 320 C. degrees water with 3 ppm H 2 , at dose rates of 7*10 -6 dpa/s and 7*10 -7 dpa/s, for 4, 24, and 72 hours. A dual-layer oxide formed on the samples, with an inner layer rich in Cr with Fe and Ni content, and an outer layer of Fe oxides. Samples were characterized with TEM (Transmission Electron Microscopy), EDS, and Raman spectroscopy to determine the effect of irradiation. Irradiated samples were found to have a thinner and more porous inner oxide which was deficient in chromium. The outer oxide was found to have significant hematite content, suggesting that irradiation led to an increase in ECP (Electro-Chemical Potential) at the oxide-solution interface, causing accelerated dissolution of the oxide under irradiation. (authors)

  8. Tensile behavior of irradiated manganese-stabilized stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Lab., TN (United States)

    1996-10-01

    Tensile tests were conducted on seven experimental, high-manganese austenitic stainless steels after irradiation up to 44 dpa in the FFTF. An Fe-20Mn-12Cr-0.25C base composition was used, to which various combinations of Ti, W, V, B, and P were added to improve strength. Nominal amounts added were 0.1% Ti, 1% W, 0.1% V, 0.005% B, and 0.03% P. Irradiation was carried out at 420, 520, and 600{degrees}C on the steels in the solution-annealed and 20% cold-worked conditions. Tensile tests were conducted at the irradiation temperature. Results were compared with type 316 SS. Neutron irradiation hardened all of the solution-annealed steels at 420, 520, and 600{degrees}C, as measured by the increase in yield stress and ultimate tensile strength. The steel to which all five elements were added to the base composition showed the least amount of hardening. It also showed a smaller loss of ductility (uniform and total elongation) than the other steels. The total and uniform elongations of this steel after irradiation at 420{degrees}C was over four times that of the other manganese-stabilized steels and 316 SS. There was much less difference in strength and ductility at the two higher irradiation temperatures, where there was considerably less hardening, and thus, less loss of ductility. In the cold-worked condition, hardening occured only after irradiation at 420{degrees}C, and there was much less difference in the properties of the steels after irradiation. At the 420{degrees}C irradiation temperature, most of the manganese-stabilized steels maintained more ductility than the 316 SS. After irradiation at 420{degrees}C, the temperature of maximum hardening, the steel to which all five of the elements were added had the best uniform elongation.

  9. Treatment of pediatric femoral shaft fractures by stainless steel and titanium elastic nail system: A randomized comparative trial

    Directory of Open Access Journals (Sweden)

    Tank Gyaneshwar

    2016-08-01

    Conclusion: Majority of paediatric femoral shaft fractures are now treated operatively by elastic stable intramedullary nails. Operative intervention results in a shorter hospital stay and has economic and social benefits over conservative treatment. The cost of stainless steel nail is one third the cost of titanium nail. However, the clinico-radiological results are not significantly different between titanium and stainless steel nails at one year follow-up as observed by our study.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  12. Depth distribution of Frank loop defects formed in ion-irradiated stainless steel and its dependence on Si addition

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Dongyue, E-mail: dychen@safety.n.t.u-tokyo.ac.jp [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan); Murakami, Kenta [The University of Tokyo, Nuclear Professional School, School of Engineering, 2-22 Shirakata-Shirane, Tokai-mura, Ibaraki 319-1188 (Japan); Dohi, Kenji; Nishida, Kenji; Soneda, Naoki [Central Research Institute of Electric Power Industry, 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Li, Zhengcao, E-mail: zcli@tsinghua.edu.cn [Tsinghua University, School of Materials Science and Engineering, Beijing 100084 (China); Liu, Li; Sekimura, Naoto [The University of Tokyo, Department of Nuclear Engineering and Management, School of Engineering, 7-3-1 Hongo Bunkyo-ku, Tokyo 113-8656 (Japan)

    2015-12-15

    Although heavy ion irradiation is a good tool to simulate neutron irradiation-induced damages in light water reactor, it produces inhomogeneous defect distribution. Such difference in defect distribution brings difficulty in comparing the microstructure evolution and mechanical degradation between neutron and heavy ion irradiation, and thus needs to be understood. Stainless steel is the typical structural material used in reactor core, and could be taken as an example to study the inhomogeneous defect depth distribution in heavy ion irradiation and its influence on the tested irradiation hardening by nano-indentation. In this work, solution annealed stainless steel model alloys are irradiated by 3 MeV Fe{sup 2+} ions at 400 °C to 3 dpa to produce Frank loops that are mainly interstitial in nature. The silicon content of the model alloys is also tuned to change point defect diffusion, so that the loop depth distribution influenced by diffusion along the irradiation beam direction could be discussed. Results show that in low Si (0% Si) and base Si (0.42% Si) samples the depth distribution of Frank loop density quite well matches the dpa profile calculated by the SRIM code, but in high Si sample (0.95% Si), the loop number density in the near-surface region is very low. One possible explanation could be Si’s role in enhancing the effective vacancy diffusivity, promoting recombination and thus suppressing interstitial Frank loops, especially in the near-surface region, where vacancies concentrate. By considering the loop depth distribution, the tested irradiation hardening is successfully explained by the Orowan model. A hardening coefficient of around 0.30 is obtained for all the three samples. This attempt in interpreting hardening results may make it easier to compare the mechanical degradation between different irradiation experiments.

  13. Intergranular stress corrosion cracking of ion irradiated 304L stainless steel in PWR environment

    International Nuclear Information System (INIS)

    Gupta, Jyoti

    2016-01-01

    IASCC is irradiation - assisted enhancement of intergranular stress corrosion cracking susceptibility of austenitic stainless steel. It is a complex degrading phenomenon which can have a significant influence on maintenance time and cost of PWRs' core internals and hence, is an issue of concern. Recent studies have proposed using ion irradiation (to be specific, proton irradiation) as an alternative of neutron irradiation to improve the current understanding of the mechanism. The objective of this study was to investigate the cracking susceptibility of irradiated SA 304L and factors contributing to cracking, using two different ion irradiations; iron and proton irradiations. Both resulted in generation of point defects in the microstructure and thereby causing hardening of the SA 304L. Material (unirradiated and iron irradiated) showed no susceptibility to intergranular cracking on subjection to SSRT with a strain rate of 5 * 10 -8 s -1 up to 4 % plastic strain in inert environment. But, irradiation (iron and proton) was found to increase intergranular cracking severity of material on subjection to SSRT in simulated PWR primary water environment at 340 C. Correlation between the cracking susceptibility and degree of localization was studied. Impact of iron irradiation on bulk oxidation of SA 304L was studied as well by conducting an oxidation test for 360 h in simulated PWR environment at 340 C. The findings of this study indicate that the intergranular cracking of 304L stainless steel in PWR environment can be studied using Fe irradiation despite its small penetration depth in material. Furthermore, it has been shown that the cracking was similar in both iron and proton irradiated samples despite different degrees of localization. Lastly, on establishing iron irradiation as a successful tool, it was used to study the impact of surface finish and strain paths on intergranular cracking susceptibility of the material. (author) [fr

  14. The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

    1994-11-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

  15. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

    1994-01-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

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

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1985-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Maziasz, P.J.

    1985-11-01

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

  18. Electron beam cladding of titanium on stainless steel plate

    International Nuclear Information System (INIS)

    Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

    1990-01-01

    Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

  19. Antibacterial effect of silver nanofilm modified stainless steel surface

    Science.gov (United States)

    Fang, F.; Kennedy, J.; Dhillon, M.; Flint, S.

    2015-03-01

    Bacteria can attach to stainless steel surfaces, resulting in the colonization of the surface known as biofilms. The release of bacteria from biofilms can cause contamination of food such as dairy products in manufacturing plants. This study aimed to modify stainless steel surfaces with silver nanofilms and to examine the antibacterial effectiveness of the modified surface. Ion implantation was applied to produce silver nanofilms on stainless steel surfaces. 35 keV Ag ions were implanted with various fluences of 1 × 1015 to 1 × 1017 ions•cm-2 at room temperature. Representative atomic force microscopy characterizations of the modified stainless steel are presented. Rutherford backscattering spectrometry spectra revealed the implanted atoms were located in the near-surface region. Both unmodified and modified stainless steel coupons were then exposed to two types of bacteria, Pseudomonas fluorescens and Streptococcus thermophilus, to determine the effect of the surface modification on bacterial attachment and biofilm development. The silver modified coupon surface fluoresced red over most of the surface area implying that most bacteria on coupon surface were dead. This study indicates that the silver nanofilm fabricated by the ion implantation method is a promising way of reducing the attachment of bacteria and delay biofilm formation.

  20. 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

  1. Effect of titanium and stainless steel posts in detection of vertical root fractures using NEWTOM VG cone beam computed tomography system

    International Nuclear Information System (INIS)

    Mohammadpour, Mahdis; Bakhshalian, Neema; Shahab, Shahriar; Sadeghi, Shaya; Ataee, Mona; Sarikhani, Soodeh

    2014-01-01

    Vertical root fracture (VRF) is a common complication in endodontically treated teeth. Considering the poor prognosis of VRF, a reliable and valid detection method is necessary. Cone beam computed tomography (CBCT) has been reported to be a reliable tool for the detection of VRF; however, the presence of metallic intracanal posts can decrease the diagnostic values of CBCT systems. This study evaluated and compared the effects of intracanal stainless steel or titanium posts on the sensitivity, specificity, and accuracy of VRF detection using a NewTom VG CBCT system. Eighty extracted single-rooted teeth were selected and sectioned at the cemento-enamel junction. The roots were divided into two groups of 40. Root fracture was induced in the test group by using an Instron machine, while the control group was kept intact. Roots were randomly embedded in acrylic blocks and radiographed with the NewTom VG, both with titanium and stainless steel posts and also without posts. Sensitivity, specificity, and accuracy values were calculated as compared to the gold standard. The sensitivity, specificity, and accuracy of VRF diagnosis were significantly lower in teeth with stainless steel and titanium posts than in those without posts. Interobserver agreement was the highest in teeth without posts, followed by stainless steel posts, and then titanium posts. Intracanal posts significantly decreased the VRF diagnostic values of CBCT. The stainless steel posts decreased the diagnostic values more than the titanium posts.

  2. Effect of titanium and stainless steel posts in detection of vertical root fractures using NEWTOM VG cone beam computed tomography system

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadpour, Mahdis [Dept. of Oral and Maxillofacial Radiology, School of Dentistry, Qazvin University of Medical Sciences, Qazvin (Iran, Islamic Republic of); Bakhshalian, Neema [Dept. of Advanced Periodontology, Ostrow School of Dentistry, University of Southern California, Los Angeles (United States); Shahab, Shahriar [Dept. of Oral and Maxillofacial Radiology, School of Dentistry, Shahed University, Tehran (Iran, Islamic Republic of); Sadeghi, Shaya; Ataee, Mona [Radmehr Oral and Maxillofacial Radiology Clinic, Ghazvin (Iran, Islamic Republic of); Sarikhani, Soodeh [Dept. of Oral and Maxillofacial Radiology, School of Dentistry, University of Golestan, Gorgan (Iran, Islamic Republic of)

    2014-06-15

    Vertical root fracture (VRF) is a common complication in endodontically treated teeth. Considering the poor prognosis of VRF, a reliable and valid detection method is necessary. Cone beam computed tomography (CBCT) has been reported to be a reliable tool for the detection of VRF; however, the presence of metallic intracanal posts can decrease the diagnostic values of CBCT systems. This study evaluated and compared the effects of intracanal stainless steel or titanium posts on the sensitivity, specificity, and accuracy of VRF detection using a NewTom VG CBCT system. Eighty extracted single-rooted teeth were selected and sectioned at the cemento-enamel junction. The roots were divided into two groups of 40. Root fracture was induced in the test group by using an Instron machine, while the control group was kept intact. Roots were randomly embedded in acrylic blocks and radiographed with the NewTom VG, both with titanium and stainless steel posts and also without posts. Sensitivity, specificity, and accuracy values were calculated as compared to the gold standard. The sensitivity, specificity, and accuracy of VRF diagnosis were significantly lower in teeth with stainless steel and titanium posts than in those without posts. Interobserver agreement was the highest in teeth without posts, followed by stainless steel posts, and then titanium posts. Intracanal posts significantly decreased the VRF diagnostic values of CBCT. The stainless steel posts decreased the diagnostic values more than the titanium posts.

  3. KSb(OH) samples previously treated with Co y - rays irradiated with neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Facetti, J F [Asuncion Nacional Univ. (Paraguay). Inst. de Ciencias

    1969-01-01

    When Ksb (OH) samples previously treated with Co y - rays or crushed are irradiated with neutrons, the yield of Sb and the annealing mechanism are apparently modified by the pretreatment. In addition it is shown that metastable species of Sb are formed under irradiation.

  4. Carbon/PEEK composite materials as an alternative for stainless steel/titanium hip prosthesis: a finite element study.

    Science.gov (United States)

    Rezaei, Farshid; Hassani, Kamran; Solhjoei, Nosratollah; Karimi, Alireza

    2015-12-01

    Total hip replacement (THR) has been ranked within the most typical surgical processes in the world. The durability of the prosthesis and loosening of prosthesis are the main concerns that mostly reported after THR surgeries. In THR, the femoral prosthesis can be fixed by either cement or cementless methods in the patient's bones. In both procedures, the stability of the prosthesis in the hosted bone has a key asset in its long-term durability and performance. This study aimed to execute a comparative finite element simulation to assess the load transfer between the prosthesis, which is made of carbon/PEEK composite and stainless steel/titanium, and the femur bone. The mechanical behavior of the cortical bone was assumed as a linear transverse isotropic while the spongy bone was modeled like a linear isotropic material. The implants were made of stainless steel (316L) and titanium alloy as they are common materials for implants. The results showed that the carbon/PEEK composites provide a flatter load transfer from the upper body to the leg compared to the stainless steel/titanium prosthesis. Furthermore, the results showed that the von Mises stress, principal stress, and the strain in the carbon/PEEK composites prosthesis were significantly lower than that made of the stainless steel/titanium. The results also imply that the carbon/PEEK composites can be applied to introduce a new optimum design for femoral prosthesis with adjustable stiffness, which can decrease the stress shielding and interface stress. These findings will help clinicians and biomedical experts to increase their knowledge about the hip replacement.

  5. Microvascular response of striated muscle to metal debris. A comparative in vivo study with titanium and stainless steel.

    Science.gov (United States)

    Kraft, C N; Diedrich, O; Burian, B; Schmitt, O; Wimmer, M A

    2003-01-01

    Wear products of metal implants are known to induce biological events which may have profound consequences for the microcirculation of skeletal muscle. Using the skinfold chamber model and intravital microscopy we assessed microcirculatory parameters in skeletal muscle after confrontation with titanium and stainless-steel wear debris, comparing the results with those of bulk materials. Implantation of stainless-steel bulk and debris led to a distinct activation of leukocytes combined with a disruption of the microvascular endothelial integrity and massive leukocyte extravasation. While animals with bulk stainless steel showed a tendency to recuperation, stainless-steel wear debris induced such severe inflammation and massive oedema that the microcirculation broke down within 24 hours after implantation. Titanium bulk caused only a transient increase in leukocyte-endothelial cell interaction within the first 120 minutes and no significant change in macromolecular leakage, leukocyte extravasation or venular diameter. Titanium wear debris produced a markedly lower inflammatory reaction than stainless-steel bulk, indicating that a general benefit of bulk versus debris could not be claimed. Depending on its constituents, wear debris is capable of eliciting acute inflammation which may result in endothelial damage and subsequent failure of microperfusion. Our results indicate that not only the bulk properties of orthopaedic implants but also the microcirculatory implications of inevitable wear debris play a pivotal role in determining the biocompatibility of an implant.

  6. Homogenous photocatalytic decontamination of prion infected stainless steel and titanium surfaces.

    Science.gov (United States)

    Berberidou, Chrysanthi; Xanthopoulos, Konstantinos; Paspaltsis, Ioannis; Lourbopoulos, Athanasios; Polyzoidou, Eleni; Sklaviadis, Theodoros; Poulios, Ioannis

    2013-01-01

    Prions are notorious for their extraordinary resistance to traditional methods of decontamination, rendering their transmission a public health risk. Iatrogenic Creutzfeldt-Jakob disease (iCJD) via contaminated surgical instruments and medical devices has been verified both experimentally and clinically. Standard methods for prion inactivation by sodium hydroxide or sodium hypochlorite have failed, in some cases, to fully remove prion infectivity, while they are often impractical for routine applications. Prion accumulation in peripheral tissues and indications of human-to-human bloodborne prion transmission, highlight the need for novel, efficient, yet user-friendly methods of prion inactivation. Here we show both in vitro and in vivo that homogenous photocatalytic oxidation, mediated by the photo-Fenton reagent, has the potential to inactivate the pathological prion isoform adsorbed on metal substrates. Photocatalytic oxidation with 224 μg mL(-1) Fe (3+), 500 μg mL(-1) h(-1) H 2O 2, UV-A for 480 min lead to 100% survival in golden Syrian hamsters after intracranial implantation of stainless steel wires infected with the 263K prion strain. Interestingly, photocatalytic treatment of 263K infected titanium wires, under the same experimental conditions, prolonged the survival interval significantly, but failed to eliminate infectivity, a result that we correlate with the increased adsorption of PrP(Sc) on titanium, in comparison to stainless steel. Our findings strongly indicate that our, user--and environmentally--friendly protocol can be safely applied to the decontamination of prion infected stainless steel surfaces.

  7. A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

    International Nuclear Information System (INIS)

    Sohn, Hee Dong

    2012-02-01

    In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

  8. A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hee Dong

    2012-02-15

    In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

  9. Titanium implants in irradiated dog mandibles

    International Nuclear Information System (INIS)

    Schweiger, J.W.

    1989-01-01

    The use of osseointegrated titanium implants has been a great benefit to selected cancer patients who otherwise would not be able to wear conventional and/or maxillofacial prostheses. Cognizant of the risk of osteoradionecrosis, we used an animal model to seek experimental evidence for successful osseointegration in bone irradiated to tumoricidal levels. Five healthy male beagle dogs received 60 gray to a previously edentulated and healed area of the right hemimandible. The left hemimandible was kept as a nonirradiated control. After 9 months, titanium implants were placed and allowed an additional 5 1/2 months to osseointegrate. At that time, block specimens were obtained, radiographed, photographed, and analyzed histologically. Although statistical significance cannot be attached to the results, osseointegration was achieved in half of the irradiated specimens

  10. 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.

  11. The role of dislocation channeling in IASCC initiation of neutron irradiated stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Stephenson, Kale J., E-mail: kalejs@umich.edu; Was, Gary S.

    2016-12-01

    This study intended to understand how dislocation channeling affects IASCC initiation using a novel four-point bend test. Stainless steels used in this study (irradiated in the BOR-60 reactor) included a commercial purity 304L alloy irradiated to 5.5, 10.2, and 47.5 dpa, and two high purity alloys, Fe−18Cr−12Ni and Fe−18Cr−25Ni, irradiated to ∼10 dpa. IASCC was enhanced by MnS inclusions, which dissolve in the NWC environment and form oxide caps, creating a crevice condition with a high propensity for crack initiation. Stress concentration at the grain boundary intersecting these sites induced crack initiation, resulting from discontinuous dislocation channels (DC). Stress to initiate IASCC decreased with dose due to earlier DC initiation. The HP Fe−18Cr−12Ni alloy had low IASCC susceptibility and the high Ni alloy did not crack. The difference was attributed to the propensity for DCs to transmit across grain boundaries, which controls stress accumulation at DC – grain boundary intersections. - Highlights: • MnS inclusions enhance susceptibility to IASCC initiation. • Local stress (typically caused by an intersecting dislocation channel) was required to cause cracking at a MnS inclusion. • The remotely applied stress at which IASCC initiation occurred decreased with increasing irradiation dose. • IASCC resistant alloys were more likely to transmit dislocation channels across grain boundaries.

  12. SB3. Experiment on secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for each of 13 different elements plus a stainless steel

    International Nuclear Information System (INIS)

    Maerker, R.E.

    1976-01-01

    The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for iron, oxygen, sodium, aluminum, copper, titanium, calcium, potassium, silicon, nickel, zinc, barium, sulfur, and a type 321 stainless steel. The gamma-ray production cross sections were binned into 0.5-MeV wide gamma-ray energy intervals. 29 tables, 1 figure

  13. Neutron irradiation of seeds 2

    Energy Technology Data Exchange (ETDEWEB)

    1968-10-01

    The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs.

  14. Neutron irradiation of seeds 2

    International Nuclear Information System (INIS)

    1968-01-01

    The irradiation of seeds with the fast neutron of research reactors has been hampered by difficulties in accurately measuring dose and in obtaining repeatable and comparable results. Co-ordinated research under an international program organized by the FAO and IAEA has already resulted in significant improvements in methods of exposing seeds in research reactors and in obtaining accurate dosimetry. This has been accomplished by the development of a standard reactor facility for the neutron irradiation of seeds and standard methods for determining fast-neutron dose and the biological response after irradiation. In this program various divisions of the IAEA and the Joint FAO/IAEA Division co-operate with a number of research institutes and reactor centres throughout the world. Results of the preliminary experiments were reported in Technical Reports Series No. 76, ''Neutron Irradiation of Seeds''. This volume contains the proceedings of a meeting of co-operators in the FAO/IAEA Neutron Seed Irradiation Program and other active scientists in this field. The meeting was held in Vienna from 11 to 15 December 1967. Refs, figs and tabs

  15. Neutron Flux Characterization of Irradiation Holes for Irradiation Test at HANARO

    Directory of Open Access Journals (Sweden)

    Yang Seong Woo

    2016-01-01

    Full Text Available The High flux Advanced Neutron Application ReactOr (HANARO is a unique research reactor in the Republic of Korea, and has been used for irradiation testing since 1998. To conduct irradiation tests for nuclear materials, the irradiation holes of CT and OR5 have been used due to a high fast-neutron flux. Because the neutron flux must be accurately calculated to evaluate the neutron fluence of irradiated material, it was conducted using MCNP. The neutron flux was measured using fluence monitor wires to verify the calculated result. Some evaluations have been conducted, however, more than 20% errors have frequently occurred at the OR irradiation hole, while a good agreement between the calculated and measured data was shown at the CT irradiation hole.

  16. Effect of stainless steel and titanium low-contact dynamic compression plate application on the vascularity and mechanical properties of cortical bone after fracture.

    Science.gov (United States)

    Jain, R; Podworny, N; Hearn, T; Anderson, G I; Schemitsch, E H

    1997-10-01

    Comparison of the effect of stainless steel and titanium low-contact dynamic compression plate application on the vascularity and mechanical properties of cortical bone after fracture. Randomized, prospective. Orthopaedic research laboratory. Ten large (greater than twenty-five kilogram) adult dogs. A short, midshaft spiral tibial fracture was created, followed by lag screw fixation and neutralization with an eight-hole, 3.5-millimeter, low-contact dynamic compression plate (LCDCP) made of either 316L stainless steel (n = five) or commercially pure titanium (n = five). After surgery, animals were kept with unrestricted weight-bearing in individual stalls for ten weeks. Cortical bone blood flow was assessed by laser Doppler flowmetry using a standard metalshafted probe (Periflux Pf303, Perimed, Jarfalla, Sweden) applied through holes in the custom-made LCDCPs at five sites. Bone blood flow was determined at four times: (a) prefracture, (b) postfracture, (c) postplating, and (d) ten weeks postplating. After the dogs were killed, the implant was removed and both the treated tibia and contralateral tibia were tested for bending stiffness and load to failure. Fracture creation decreased cortical perfusion in both groups at the fracture site (p = 0.02). The application of neither stainless steel nor titanium LCDCPs further decreased cortical bone blood flow after fracture creation. However, at ten weeks postplating, cortical perfusion significantly increased compared with acute postplating levels in the stainless steel (p = 0.003) and titanium (p = 0.001) groups. Cortical bone blood flow ten weeks postplating was not significantly different between the titanium group and the stainless steel group. Biomechanical tests performed on the tibiae with the plates removed did not reveal any differences in bending stiffness nor load required to cause failure between the two groups. Both titanium and stainless steel LCDCPs were equally effective in allowing revascularization, and

  17. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Qiaofeng [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Su, Qing [Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Wang, Fei [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Zhang, Chenfei; Lu, Yongfeng [Department of Electrical Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nastasi, Michael [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Energy Sciences Research, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Cui, Bai, E-mail: bcui3@unl.edu [Department of Mechanical & Materials Engineering, University of Nebraska–Lincoln, Lincoln, NE 68588 (United States); Nebraska Center for Materials and Nanoscience, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States)

    2017-06-15

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments. - Highlights: •Laser shock peening generates a dislocation network, stacking faults and deformation twins in stainless steels. •Dislocations and incoherent twin boundaries serve as effective sinks for the annihilation of irradiation defects. •Incoherent twin boundaries remain as stable and effective defect sinks at 300 °C.

  18. Effect of irradiation temperature on microstructural changes in self-ion irradiated austenitic stainless steel

    Science.gov (United States)

    Jin, Hyung-Ha; Ko, Eunsol; Lim, Sangyeob; Kwon, Junhyun; Shin, Chansun

    2017-09-01

    We investigated the microstructural and hardness changes in austenitic stainless steel after Fe ion irradiation at 400, 300, and 200 °C using transmission electron microscopy (TEM) and nanoindentation. The size of the Frank loops increased and the density decreased with increasing irradiation temperature. Radiation-induced segregation (RIS) was detected across high-angle grain boundaries, and the degree of RIS increases with increasing irradiation temperature. Ni-Si clusters were observed using high-resolution TEM in the sample irradiated at 400 °C. The results of this work are compared with the literature data of self-ion and proton irradiation at comparable temperatures and damage levels on stainless steels with a similar material composition with this study. Despite the differences in dose rate, alloy composition and incident ion energy, the irradiation temperature dependence of RIS and the size and density of radiation defects followed the same trends, and were very comparable in magnitude.

  19. Estimates of time-dependent fatigue behavior of type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems

    International Nuclear Information System (INIS)

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

    1979-01-01

    Cyclic lives obtained from strain-controlled fatigue tests at 593 0 C of specimens irraidated in the experimental breeder reactor II (EBR-II) to a fluence of 1 to 2.63 x 10 26 neutrons (n)/m 2 E > 0.1 MeV) were compared with predictions based on the method of strain-range partitioning. It was demonstrated that, when appropriate tensile and creep-rupture ductilities were employed, reasonably good estimates of the influence of hold periods and irradiation damage on the fully reversed fatigue life of Type 316 stainless steel could be made. After applicability of this method was demonstrated, ductility values for 20% cold-worked Type 316 stainless steel specimens irradiated in a mixed-spectrum fission reactor were used to estimate fusion reactor first-wall lifetime. The ductility values used were from irradations that simulate the environment of the first wall of a fusion reactor. Neutron wall loadins ranging from 2 to 5 MW/m 2 were used. Results, although conjectural because of the many assumptions, tended to show that 20% cold-worked Type 316 stainless steel could be used as a first-wall material meeting a 7.5 go 8.5 MW-year/m 2 lifetime goal provided the neutron wall loading does not exceed more than about 2 MW/m 2 . These results were obtained for an air environment, ant it is expected that the actual vacuum environment will extend lifetime beyond 10 MW-year/m 2

  20. Influence of laser shock peening on irradiation defects in austenitic stainless steels

    Science.gov (United States)

    Lu, Qiaofeng; Su, Qing; Wang, Fei; Zhang, Chenfei; Lu, Yongfeng; Nastasi, Michael; Cui, Bai

    2017-06-01

    The laser shock peening process can generate a dislocation network, stacking faults, and deformation twins in the near surface of austenitic stainless steels by the interaction of laser-driven shock waves with metals. In-situ transmission electron microscopy (TEM) irradiation studies suggest that these dislocations and incoherent twin boundaries can serve as effective sinks for the annihilation of irradiation defects. As a result, the irradiation resistance is improved as the density of irradiation defects in laser-peened stainless steels is much lower than that in untreated steels. After heating to 300 °C, a portion of the dislocations and stacking faults are annealed out while the deformation twins remain stable, which still provides improved irradiation resistance. These findings have important implications on the role of laser shock peening on the lifetime extension of austenitic stainless steel components in nuclear reactor environments.

  1. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, A.V., E-mail: Alexey.V.Subbotin@gmail.com [Scientific and Production Complex Atomtechnoprom, Moscow 119180 (Russian Federation); Panyukov, S.V., E-mail: panyukov@lpi.ru [PN Lebedev Physics Institute, Russian Academy of Sciences, Moscow 117924 (Russian Federation)

    2016-08-15

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  2. Effect of Neutron Irradiation on Beam-Column Interaction of Reinforced Concrete

    International Nuclear Information System (INIS)

    Kwon, Tae-Hyun; Park, Jiho; Kim, Jun Yeon; Kim, HyungTae; Park, Kyoungsoo; Kim, Sang-Ho

    2015-01-01

    Age-related effects on such RC structures have been extensively studied in detail. However, the effect of neutron irradiation requires further studies from its limited database. Most of RC structures have been regarded as sound as the neutron fluence below 1.0x10 19 n/cm 2 . The reduction of strength is not considered in a periodic inspection program at aging NPPs. However, RC structures, such as biological shields and supports for a reactor vessel, could be exposed to see the critical level of neutron fluence at years of operation. In this regard, beam-column interaction of a typical RC member is numerically investigated as a result of neutron irradiation. The effect of neutron irradiation on beam-column interaction is evaluated. ACI318 requires the strength reduction factor, ϕ=0.70, for the compression controlled area and the higher up to 0.9 as the tensile strain in steel reinforcement goes higher. This concept works well with this example. However, this does not take into account the energy dissipation capacity of the member but it only expresses the ultimate strength. Therefore, the current strength evaluation concept may be misleading when the material behavior of steel reinforcement becomes brittle due to the neutron irradiation. In such case, even for the transient and tension controlled area, the strength reduction factor needs to be modified to account for the potential ductility loss

  3. Neutron metrology in the HFR. Steel irradiation R139-699 (SINAS)

    International Nuclear Information System (INIS)

    Baard, J.H.; Paardekooper, A.

    1996-06-01

    The aim of the irradiation of experiment R139-699 was the testing of austenitic stainless steel type AISI-316 TIG. This report presents the final metrology results obtained from activation monitors in the specimen holder, coded as R139-699. Data about the helium production as well as the number of displacements per atom are also included. The irradiation circumstances for this experiment, carried out in a TRIO type capsule in HFR position F2, are as close as possible relevant for the candidate materials which will be used for the first wall of the NET (Next European Torus). The main results of the thermal and fast neutron fluence measurements are presented in tables 2 and 3 as well as in the figure 2. (orig.)

  4. Aging and Embrittlement of High Fluence Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Was, gary; Jiao, Zhijie; der ven, Anton Van; Bruemmer, Stephen; Edwards, Dan

    2012-12-31

    Irradiation of austenitic stainless steels results in the formation of dislocation loops, stacking fault tetrahedral, Ni-Si clusters and radiation-induced segregation (RIS). Of these features, it is the formation of precipitates which is most likely to impact the mechanical integrity at high dose. Unlike dislocation loops and RIS, precipitates exhibit an incubation period that can extend from 10 to 46 dpa, above which the cluster composition changes and a separate phase, (G-phase) forms. Both neutron and heavy ion irradiation showed that these clusters develop slowly and continue to evolve beyond 100 dpa. Overall, this work shows that the irradiated microstructure features produced by heavy ion irradiation are remarkably comparable in nature to those produced by neutron irradiation at much lower dose rates. The use of a temperature shift to account for the higher damage rate in heavy ion irradiation results in a fairly good match in the dislocation loop microstructure and the precipitate microstructure in austenitic stainless steels. Both irradiations also show segregation of the same elements and in the same directions, but to achieve comparable magnitudes, heavy ion irradiation must be conducted at a much higher temperature than that which produces a match with loops and precipitates. First-principles modeling has confirmed that the formation of Ni-Si precipitates under irradiation is likely caused by supersaturation of solute to defect sinks caused by highly correlated diffusion of Ni and Si. Thus, the formation and evolution of Ni-Si precipitates at high dose in austenitic stainless steels containing Si is inevitable.

  5. Neutron metrology in the HFR. Steel irradiation R139-805 (SINAS)

    International Nuclear Information System (INIS)

    Baard, J.H.

    1996-10-01

    The experiment R139-80 is part of a material program to test austenitic stainless steel of different types and has been irradiated in the HFR Petten. This report presents the final metrology results obtained from activation monitors in the specimen holder, coded as R139-805. Data about the helium production as well as the number of displacements per atom are included. The irradiation circumstances for this experiment, carried out in a TRIO type capsule in HFR position F2, represent the conditions at the first wall of NET (Next European Torus). The aim of this irradiation for specimen holder R139-805 was to reach a damage level of 2.4 dpa at a temperature of 325 C. However, the specimens have been irradiated up to a damage level of 1.7-2.0 dpa. The main results of the thermal and fast neutron fluence measurements are presented in tables 2 and 3 as well as in the figure 2. (orig.)

  6. Effects of Admixed Titanium on Densification of 316L Stainless Steel Powder during Sintering

    Directory of Open Access Journals (Sweden)

    Aslam Muhammad

    2014-07-01

    Full Text Available Effects of admixed titanium on powder water atomized (PWA and powder gas atomized (PGA 316L stainless steel (SS have been investigated in terms of densification. PGA and PWA powders, having different shapes and sizes, were cold pressed and sintered in argon atmosphere at 1300°C. The admixed titanium compacts of PGA and PWA have shown significant effect on densification through formation of intermetallic compound and reducing porosity during sintering process. PWA, having particle size 8 μm, blended with 1wt% titanium has exhibited higher sintered density and shrinkage as compared to gas atomized powder compacts. Improved densification of titanium blended PGA and PWA 316L SS at sintering temperature 1300°C is probably due to enhanced diffusion kinetics resulting from stresses induced by concentration gradient in powder compacts.

  7. 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

  8. The characteristics of corrosion, radiation degradation and dissolution of titanium alloys

    International Nuclear Information System (INIS)

    Sung, K. W.; Na, J. W.; Choi, B. S.; Lee, D. J.; Chang, M. H.

    2001-12-01

    In order to establish the technical bases of water chemistry design requirement related titanium alloys, we investigated the characteristics of corrosion, activation, radiation degradation, radiation hydrogen embrittlement of titanium alloys and dissolution of titanium dioxide. Titanium alloys generally have high corrosion resistance. Corrosion product release from PT-7M and PT-3V titanium alloy surface for 18 months of operation is negligible, and the corrosion penetration for about 30 years is about 1 μm, while the corrosion rates is not higher than one third of that of austenitic steel. Titanium only converts into Sc-46 with 85 day halflife after neutron irradiation, and its radioactivity is not higher than one thousandth of that produced from nickel. Therefore, under the condition without any neutron irradiation, the radiation damage of titanium alloys would have no problem. Titanium dioxide, that protects the metals from the corrosion, has retrograde solubility in neutral solutions. It does not form any complexes with ligands such as ammonia, but Ti(IV) gets more stable by complexing with water molecules. In conclusion, it is estimated that titanium alloys such as PT-7M would be applicable to steam generator materials

  9. Potential for photocatalytic degradation of the potassic diclofenac using scandium and silver modified titanium dioxide thin films

    International Nuclear Information System (INIS)

    Ciola, R.A.; Oliveira, C.T.; Lopes, S.A.; Cavalheiro, A.A.

    2011-01-01

    The potential for photocatalytic degradation of the potassic diclofenac drug was investigated using titanium dioxide thin films modified with two modifier types, scandium and silver, both prepared by Sol-Gel method. It was demonstrated by UVVis spectroscopy analysis of the solutions containing the drug, under UV-A light irradiation that the degradation efficiency of the titanium dioxide photocatalyst is dependent of the semiconductor nature and that the scandium accelerates the first step of the degradation when compared to the silver. This result seems to be related to the redox potential of the electron-hole pair, once the scandium modifying sample generates a p type semiconductor that reduces the band gap. The extra holes attract more strongly the chorine ion present in diclofenac and leading to the releasing more easily. However, after the first byproducts degradation the following steps are not facilitated, making the silver modifying more advantageous. (author)

  10. Grafting of HEMA onto dopamine coated stainless steel by 60Co-γ irradiation method

    International Nuclear Information System (INIS)

    Jin, Wanqin; Yang, Liming; Yang, Wei; Chen, Bin; Chen, Jie

    2014-01-01

    A novel method for grafting of 2-hydroxyethyl methacrylate (HEMA) onto the surface of stainless steel (SS) was explored by using 60 Co-γ irradiation. The surface of SS was modified by coating of dopamine before radiation grafting. The grafting reaction was performed in a simultaneous irradiation condition. The chemical structures change of the surface before and after grafting was demonstrated by Fourier transform infrared (FTIR) spectrometer. The hydrophilicity of the samples was determined by water contact angle measurement in the comparison of the stainless steel in the conditions of pristine, dopamine coated and HEMA grafted. Surface morphology of the samples was characterized by atomic force microscope (AFM) and scanning electron microscope (SEM). The corrosion resistance properties of the samples were evaluated by Tafel polarization curve. The hemocompatibility of the samples were tested by platelet adhesion assay. - Highlights: • Poly-HEMA was grafted onto the surface of SS by 60 Co-γ-ray irradiation. • Pristine SS was coated by dopamine to form a dense poly-dopamine film before radiation grafting. • The biocompatibility and hydrophility of SS were improved after the grafting of HEMA

  11. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  12. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  13. Neutron irradiation creep in stainless steel alloys

    Energy Technology Data Exchange (ETDEWEB)

    Schuele, Wolfgang (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy)); Hausen, Hermann (Commission of the European Union, Institute for Advanced Materials, I-21020 Ispra (Vatican City State, Holy See) (Italy))

    1994-09-01

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

  14. Comparative in vitro biocompatibility of nickel-titanium, pure nickel, pure titanium, and stainless steel: genotoxicity and atomic absorption evaluation.

    Science.gov (United States)

    Assad, M; Lemieux, N; Rivard, C H; Yahia, L H

    1999-01-01

    The genotoxicity level of nickel-titanium (NiTi) was compared to that of its pure constituents, pure nickel (Ni) and pure titanium (Ti) powders, and also to 316L stainless steel (316L SS) as clinical reference material. In order to do so, a dynamic in vitro semiphysiological extraction was performed with all metals using agitation and ISO requirements. Peripheral blood lymphocytes were then cultured in the presence of all material extracts, and their comparative genotoxicity levels were assessed using electron microscopy-in situ end-labeling (EM-ISEL) coupled to immunogold staining. Cellular chromatin exposition to pure Ni and 316L SS demonstrated a significantly stronger gold binding than exposition to NiTi, pure Ti, or the untreated control. In parallel, graphite furnace atomic absorption spectrophotometry (AAS) was also performed on all extraction media. The release of Ni atoms took the following decreasing distribution for the different resulting semiphysiological solutions: pure Ni, 316L SS, NiTi, Ti, and controls. Ti elements were detected after elution of pure titanium only. Both pure titanium and nickel-titanium specimens obtained a relative in vitro biocompatibility. Therefore, this quantitative in vitro study provides optimistic results for the eventual use of nickel-titanium alloys as surgical implant materials.

  15. Anodized titanium and stainless steel in contact with CFRP: an electrochemical approach considering galvanic corrosion.

    Science.gov (United States)

    Mueller, Yves; Tognini, Roger; Mayer, Joerg; Virtanen, Sannakaisa

    2007-09-15

    The combination of different materials in an implant gives the opportunity to better fulfill the requirements that are needed to improve the healing process. However, using different materials increases the risk of galvanic coupling corrosion. In this study, coupling effects of gold-anodized titanium, stainless steel for biomedical applications, carbon fiber reinforced polyetheretherketone (CFRP), and CFRP containing tantalum fibers are investigated electrochemically and by long-term immersion experiments in simulated body fluid (SBF). Potentiodynamic polarization experiments (i/E curves) and electrochemical impedance spectroscopy (EIS) of the separated materials showed a passive behavior of the metallic samples. Anodized titanium showed no corrosion attacks, whereas stainless steel is highly susceptibility for localized corrosion. On the other side, an active dissolution behavior of both of the CFRPs in the given environment could be determined, leading to delaminating of the carbon fibers from the matrix. Long-term immersion experiments were carried out using a set-up especially developed to simulate coupling conditions of a point contact fixator system (PC-Fix) in a biological environment. Electrochemical data were acquired in situ during the whole immersion time. The results of the immersion experiments correlate with the findings of the electrochemical investigation. Localized corrosion attacks were found on stainless steel, whereas anodized titanium showed no corrosion attacks. No significant differences between the two CFRP types could be found. Galvanic coupling corrosion in combination with crevice conditions and possible corrosion mechanisms are discussed. Copyright 2007 Wiley Periodicals, Inc.

  16. Neutron irradiation damage in transition metal carbides

    International Nuclear Information System (INIS)

    Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

    1991-01-01

    Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

  17. Modeling of cavity swelling-induced embrittlement in irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Han, X.

    2012-01-01

    During long-time neutron irradiation occurred in Pressurized Water Reactors (PWRs), significant changes of the mechanical behavior of materials used in reactor core internals (made of 300 series austenitic stainless steels) are observed, including irradiation induced hardening and softening, loss of ductility and toughness. So far, much effect has been made to identify radiation effects on material microstructure evolution (dislocations, Frank loops, cavities, segregation, etc.). The irradiation-induced cavity swelling, considered as a potential factor limiting the reactor lifetime, could change the mechanical properties of materials (plasticity, toughness, etc.), even lead to a structure distortion because of the dimensional modifications between different components. The principal aim of the present PhD work is to study qualitatively the influence of cavity swelling on the mechanical behaviors of irradiated materials. A micromechanical constitutive model based on dislocation and irradiation defect (Frank loops) density evolution has been developed and implemented into ZeBuLoN and Cast3M finite element codes to adapt the large deformation framework. 3D FE analysis is performed to compute the mechanical properties of a polycrystalline aggregate. Furthermore, homogenization technique is applied to develop a Gurson-type model. Unit cell simulations are used to study the mechanical behavior of porous single crystals, by accounting for various effects of stress triaxiality, of void volume fraction and of crystallographic orientation, in order to study void effect on the irradiated material plasticity and roughness at polycrystalline scale. (author) [fr

  18. Effect of phosphorus on the swelling and precipitation behavior of austenitic stainless steels during irradiation

    International Nuclear Information System (INIS)

    Lee, E.H.; Mansur, L.K.; Rowcliffe, A.F.

    1983-01-01

    It has been observed that increasing the volume fraction of the needle-shaped iron phosphide phase in austenitic stainless steels tends to inhibit void swelling during neutron irradiation. An earlier analysis showed that this effect could not be accounted for in terms of enhanced point defect recombination at particle-matrix interfaces. The behavior of the iron phosphide phase has been further examined using dual ion beam irradiations. It was found that the particle-matrix interface serves as a site for the nucleation of a very fine dispersion of helium bubbles. It is thought that since a high number density of cavities lowers the number of helium atoms per cavity, the irradiation time for the cavities to accumulate the critical number of gas atoms for bias-driven growth is correspondingly increased. Although the phosphide phase nucleates rapidly, it eventually undergoes dissolution if either the G or Laves phase develops with increasing dose

  19. 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

  20. Thermal neutron capture cross section of chromium, vanadium, titanium and nickel isotopes

    International Nuclear Information System (INIS)

    Venturini, L.; Pecequilo, B.R.S.

    1990-04-01

    The thermal neutron cross section of chromium, vanadium, titanium and nickel can be determined by measuring the pair spectrum of prompt gamma-rays emitted targets of these elements are irradiated by a thermal neutron beam. Such measurements were carried out by irradiating the natural element mixed with a nitrogen standard (melamine) in the tangential beam hole of the IEA-R1 research reactor. The pair spectrometer efficiency calibration curve in the 1.5 to 11 MeV energy range was performed with a melamine plus ammonium chloride mixed target. The cross section was calculated for the most prominent gamma transitions of each isotope, using nitrogen as standard and averaged over the obtained values. The resulting mean cross sections are as follows: (13.4 ± 0.7)b for 50 Cr, (0.79 ± 0,02)b for 52 Cr, (18.1 ± 0,7)b for 53 Cr, (4.9 ± 0.2)b for 51 V, (8.4 ± 0.1)b for 48 Ti, (4.41 ± 0.08)b 58 Ni, (2.54 ± 0.07)b for 60 Ni, (15.2 ± 0.5)b for 62 Ni and (1.6 ± 0.1) for 64 Ni. (author) [pt

  1. Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

    International Nuclear Information System (INIS)

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

    2014-01-01

    The behavior of D retentions for Fe 2+ irradiated tungsten with the damage of 0.025-3 dpa was compared with that for neutron irradiated tungsten with 0.025 dpa. The D 2 TDS spectra for Fe 2+ irradiated tungsten consisted of two desorption stages at 450 K and 550 K although that for neutron irradiated tungsten was composed of three stages and addition desorption stage was found around 750 K. The desorption rate of major desorption stage at 550 K increased as the number of dpa by Fe 2+ irradiation increased. In addition, the first desorption stage at 450 K was only found for the damaged samples, indicating that the second stage would be based on intrinsic defects or vacancy produced by Fe 2+ irradiation and the first stage should be the accumulation of D in mono vacancy leading to the lower activation energy, where the dislocation loop and vacancy was produced. The third one was only found for the neutron irradiation, showing the D trapping by void or vacancy cluster and the diffusion effect is also contributed due to high FWHM of TDS spectrum. It can be said that the D 2 TDS spectra for Fe 2+ -irradiated tungsten could not represent that for neutron-irradiated one, showing that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten has a difference from that for ion-irradiated one. (author)

  2. On the Possibility of Laser Cladding for 304 Stainless Steel using Commercially Pure Titanium

    OpenAIRE

    Hashem F. El-Labban; Essam R.I. Mahmoud

    2014-01-01

    This work is an attempt to study the possibility of cladding the 304 stainless steel with commercially pure titanium powder using YAG fiber laser. The treatments were carried out at powers of 2800, 2400 and 2000 W and travelling speeds of 4 and 8 mm/s. In the titanium side, acicular α' martensite structure was produced. At low travelling speed (4 mm/s), coarse intermetallic phases (FeTi and Fe2Ti) were formed, and decohesion were resulted at the interface between the cladding layer and the su...

  3. Crystalline hydroxyapatite coatings synthesized under hydrothermal conditions on modified titanium substrates

    International Nuclear Information System (INIS)

    Suchanek, Katarzyna; Bartkowiak, Amanda; Gdowik, Agnieszka; Perzanowski, Marcin; Kąc, Sławomir; Szaraniec, Barbara; Suchanek, Mateusz; Marszałek, Marta

    2015-01-01

    Hydroxyapatite coatings were successfully produced on modified titanium substrates via hydrothermal synthesis in a Ca(EDTA) 2− and (NH 4 ) 2 HPO 4 solution. The morphology of modified titanium substrates as well as hydroxyapatite coatings was studied using scanning electron microcopy and phase identification by X-ray diffraction, and Raman and FTIR spectroscopy. The results show that the nucleation and growth of hydroxyapatite needle-like crystals with hexagonal symmetry occurred only on titanium substrates both chemically and thermally treated. No hydroxyapatite phase was detected on only acid etched Ti metal. This finding demonstrates that only a particular titanium surface treatment can effectively induce the apatite nucleation under hydrothermal conditions. - Highlights: • Bioactivation of titanium substrate by chemical and heat treatments • Precipitation of hydroxyapatite on modified titanium plates • Hydrothermal crystallization of hydroxyapatite by chelate decomposition method

  4. Crystalline hydroxyapatite coatings synthesized under hydrothermal conditions on modified titanium substrates

    Energy Technology Data Exchange (ETDEWEB)

    Suchanek, Katarzyna, E-mail: Katarzyna.Suchanek@ifj.edu.pl [The Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego Street 152, 31-342 Krakow (Poland); Bartkowiak, Amanda [The Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego Street 152, 31-342 Krakow (Poland); Gdowik, Agnieszka [Faculty of Physics and Applied Computer Science, AGH University of Science and Technology, Mickiewicza 30, 30-059 Krakow (Poland); Perzanowski, Marcin [The Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego Street 152, 31-342 Krakow (Poland); Kąc, Sławomir [Faculty of Metals Engineering and Industrial Computer Science, AGH University of Science and Technology, Mickiewica 30, 30-059 Krakow (Poland); Szaraniec, Barbara [Department of Biomaterials, AGH University of Science and Technology, Mickiewicza 30, 30-059 Krakow (Poland); Suchanek, Mateusz [Department of Chemistry and Physics, University of Agriculture in Krakow, Mickiewicza 21, 31-120 Krakow (Poland); Marszałek, Marta [The Henryk Niewodniczański Institute of Nuclear Physics, Polish Academy of Sciences, Radzikowskiego Street 152, 31-342 Krakow (Poland)

    2015-06-01

    Hydroxyapatite coatings were successfully produced on modified titanium substrates via hydrothermal synthesis in a Ca(EDTA){sup 2−} and (NH{sub 4}){sub 2}HPO{sub 4} solution. The morphology of modified titanium substrates as well as hydroxyapatite coatings was studied using scanning electron microcopy and phase identification by X-ray diffraction, and Raman and FTIR spectroscopy. The results show that the nucleation and growth of hydroxyapatite needle-like crystals with hexagonal symmetry occurred only on titanium substrates both chemically and thermally treated. No hydroxyapatite phase was detected on only acid etched Ti metal. This finding demonstrates that only a particular titanium surface treatment can effectively induce the apatite nucleation under hydrothermal conditions. - Highlights: • Bioactivation of titanium substrate by chemical and heat treatments • Precipitation of hydroxyapatite on modified titanium plates • Hydrothermal crystallization of hydroxyapatite by chelate decomposition method.

  5. Evaluation of neutron irradiation fields for BNCT by using absorbed dose in a phantom

    International Nuclear Information System (INIS)

    Aizawa, O.

    1993-01-01

    In a previous paper, the author defined the open-quotes irradiation timeclose quotes as the time of irradiation in which the maximum open-quotes total background doseclose quotes becomes 2,500 RBE-cGy. In this paper, he has modified the definition a little as the time of irradiation in which the maximum open-quotes lμg/g B-10 doseclose quotes becomes 3,000 RBE-cGy, because he assumed that normal tissue contained 1μg/g B-10. Moreover, he has modified the dose criteria for BNCT as follows: The open-quotes eye doseclose quotes, open-quotes total body doseclose quotes and open-quotes except-head doseclose quotes should be less that 200, 100 and 50 RBE-cGy, respectively. He has added one more criterion for BNCT that the thermal neutron fluence at the tumor position should be over 2.5x10 12 n/cm 2 at the open-quotes irradiation timeclose quotes. The distance from the core side to the irradiation port in the open-quotes old configurationclose quotes of the Musashi reactor (TRIGA-II, 100kW) was 160 cm. He is now planning to design an eccentric core and to move the reactor core nearer to the irradiation port, distance between the core side and the irradiation port to be 140, 130 and 120cm. The other assumptions used in this paper are as follows: (1) The B-10 concentrations in tumor are 30 and/or 10μg/g. (2) The depth of the tumor is 5.0 cm to 5.5 cm from the surface. (3) The RBE values used are 1.0 for all gamma rays and 2.3 for B 10 (n,α) reaction products. (4) The RBE values for neutrons are the following three cases: the first case is using 1.6 for all neutrons; the second one is using 3.2 for non-thermal neutrons and 1.6 for thermal neutrons; the third case is using 4.8 for fast neutrons, 3.2 for faster epithermal and epithermal neutrons, and 1.6 for thermal neutrons

  6. In vivo evaluation of immediately loaded stainless steel and titanium orthodontic screws in a growing bone.

    Directory of Open Access Journals (Sweden)

    Kerstin Gritsch

    Full Text Available The present work intends to evaluate the use of immediate loaded orthodontic screws in a growing model, and to study the specific bone response. Thirty-two screws (half of stainless steel and half of titanium were inserted in the alveolar bone of 8 growing pigs. The devices were immediately loaded with a 100 g orthodontic force. Two loading periods were assessed: 4 and 12 weeks. Both systems of screws were clinically assessed. Histological observations and histomorphometric analysis evaluated the percent of "bone-to-implant contact" and static and dynamic bone parameters in the vicinity of the devices (test zone and in a bone area located 1.5 cm posterior to the devices (control zone. Both systems exhibit similar responses for the survival rate; 87.5% and 81.3% for stainless steel and titanium respectively (p = 0.64; 4-week period, and 62.5% and 50.0% for stainless steel and titanium respectively (p = 0.09; 12-week period. No significant differences between the devices were found regarding the percent of "bone-to-implant contact" (p = 0.1 or the static and dynamic bone parameters. However, the 5% threshold of "bone-to-implant contact" was obtained after 4 weeks with the stainless steel devices, leading to increased survival rate values. Bone in the vicinity of the miniscrew implants showed evidence of a significant increase in bone trabecular thickness when compared to bone in the control zone (p = 0.05. In our study, it is likely that increased trabecular thickness is a way for low density bone to respond to the stress induced by loading.

  7. In vivo evaluation of immediately loaded stainless steel and titanium orthodontic screws in a growing bone.

    Science.gov (United States)

    Gritsch, Kerstin; Laroche, Norbert; Bonnet, Jeanne-Marie; Exbrayat, Patrick; Morgon, Laurent; Rabilloud, Muriel; Grosgogeat, Brigitte

    2013-01-01

    The present work intends to evaluate the use of immediate loaded orthodontic screws in a growing model, and to study the specific bone response. Thirty-two screws (half of stainless steel and half of titanium) were inserted in the alveolar bone of 8 growing pigs. The devices were immediately loaded with a 100 g orthodontic force. Two loading periods were assessed: 4 and 12 weeks. Both systems of screws were clinically assessed. Histological observations and histomorphometric analysis evaluated the percent of "bone-to-implant contact" and static and dynamic bone parameters in the vicinity of the devices (test zone) and in a bone area located 1.5 cm posterior to the devices (control zone). Both systems exhibit similar responses for the survival rate; 87.5% and 81.3% for stainless steel and titanium respectively (p = 0.64; 4-week period), and 62.5% and 50.0% for stainless steel and titanium respectively (p = 0.09; 12-week period). No significant differences between the devices were found regarding the percent of "bone-to-implant contact" (p = 0.1) or the static and dynamic bone parameters. However, the 5% threshold of "bone-to-implant contact" was obtained after 4 weeks with the stainless steel devices, leading to increased survival rate values. Bone in the vicinity of the miniscrew implants showed evidence of a significant increase in bone trabecular thickness when compared to bone in the control zone (p = 0.05). In our study, it is likely that increased trabecular thickness is a way for low density bone to respond to the stress induced by loading.

  8. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  9. Neutron metrology in the HFR. Steel irradiation R139-697/698

    International Nuclear Information System (INIS)

    Hoving, A.H.; Voorbraak, W.P.

    1996-06-01

    The experiment R139-697/698, for testing the austenitic stainless steel type AISI-316 TIG, has been irradiated in the HFR Petten. This report presents the final metrology results obtained from activation monitors in the specimen holders, coded as R139/697 and R139/698. Data about the helium production as well as the number of displacements per atom are also included. The irradiation conditions for this experiment, carried out in a TRIO type capsule in HFR position F2, are as close as possible relevant for the candidate materials which will be used for the first wall of the NET (Next European Torus). The main results of the thermal and fast neutron fluence mesurements are presented in table 2 and 3 as well as in figures 3 and 4. (orig.)

  10. Does surface anodisation of titanium implants change osseointegration and make their extraction from bone any easier?

    Science.gov (United States)

    Langhoff, J D; Mayer, J; Faber, L; Kaestner, S B; Guibert, G; Zlinszky, K; Auer, J A; von Rechenberg, B

    2008-01-01

    Titanium implants have a tendency for high bone-implant bonding, and, in comparison to stainless steel implants are more difficult to remove. The current study was carried out to evaluate, i) the release strength of three selected anodized titanium surfaces with increased nanohardness and low roughness, and ii) bone-implant bonding in vivo. These modified surfaces were intended to give improved anchorage while facilitating easier removal of temporary implants. The new surfaces were referenced to a stainless steel implant and a standard titanium implant surface (TiMAX). In a sheep limb model, healing period was 3 months. Bone-implant bonding was evaluated either biomechanically or histologically. The new surface anodized screws demonstrated similar or slightly higher bone-implant-contact (BIC) and torque release forces than the titanium reference. The BIC of the stainless steel implants was significant lower than two of the anodized surfaces (p = 0.04), but differences between stainless steel and all titanium implants in torque release forces were not significant (p = 0.06). The new anodized titanium surfaces showed good bone-implant bonding despite a smooth surface and increased nanohardness. However, they failed to facilitate implant removal at 3 months.

  11. Corrosion resistance of a magnetic stainless steel ion-plated with titanium nitride.

    Science.gov (United States)

    Hai, K; Sawase, T; Matsumura, H; Atsuta, M; Baba, K; Hatada, R

    2000-04-01

    This in vitro study evaluated the corrosion resistance of a titanium nitride (TiN) ion-plated magnetic stainless steel (447J1) for the purpose of applying a magnetic attachment system to implant-supported prostheses made of titanium. The surface hardness of the TiN ion-plated 447J1 alloy with varying TiN thickness was determined prior to the corrosion testing, and 2 micrometers thickness was confirmed to be appropriate. Ions released from the 447J1 alloy, TiN ion-plated 447J1 alloy, and titanium into a 2% lactic acid aqueous solution and 0.1 mol/L phosphate buffered saline (PBS) were determined by means of an inductively coupled plasma atomic emission spectroscopy (ICP-AES). Long-term corrosion behaviour was evaluated using a multisweep cyclic voltammetry. The ICP-AES results revealed that the 447J1 alloy released ferric ions into both media, and that the amount of released ions increased when the alloy was coupled with titanium. Although both titanium and the TiN-plated 447J1 alloy released titanium ions into lactic acid solution, ferric and chromium ions were not released from the alloy specimen for all conditions. Cyclic voltamograms indicated that the long-term corrosion resistance of the 447J1 alloy was considerably improved by ion-plating with TiN.

  12. Interfacial microstructure and mechanical properties of diffusion-bonded titanium-stainless steel joints using a nickel interlayer

    International Nuclear Information System (INIS)

    Kundu, S.; Chatterjee, S.

    2006-01-01

    Diffusion bonding was carried out between commercially pure titanium and 304 stainless steel using nickel interlayer in the temperature range of 800-950 deg. C for 3.6 ks under 3 MPa load in vacuum. The transition joints thus formed were characterized in optical and scanning electron microscopes. TiNi 3 , TiNi and Ti 2 Ni are formed at the nickel-titanium (Ni-Ti) interface; whereas, stainless steel-nickel (SS-Ni) interface is free from intermetallic compounds up to 900 deg. C processing temperatures. At 950 deg. C, Ni-Ti interface exhibits the presence of β-Ti discrete islands in the matrix of Ti 2 Ni and the phase mixture of λ + χ + α-Fe, λ + α-Fe, λ + FeTi + β-Ti and FeTi + β-Ti occurs at the stainless steel-nickel interface. Nickel is able to inhibit the diffusion of Ti to stainless steel side up to 900 deg. C temperature; however, becomes unable to restrict the migration of Ti to stainless steel at 950 deg. C. Bond strength was also evaluated and maximum tensile strength of ∼302 MPa and shear strength of ∼219 MPa were obtained for the diffusion couple processed at 900 deg. C temperature due to better contact of the mating surfaces and failure takes place at the Ni-Ti interface. At higher joining temperature, the formation of Fe-Ti bases intermetallics reduces the bond strength and failure occurs at the SS-Ni interface

  13. INTERWELD - European project to determine irradiation induced material changes in the heat affected zones of austenitic stainless steel welds that influence the stress corrosion behaviour in high-temperature water

    International Nuclear Information System (INIS)

    Roth, A.; Schaaf, Bob van der; Castano, M.L.; Ohms, C.; Gavillet, D.; Dyck, S. van

    2003-01-01

    PWR and BWR RPV internals have experienced stress corrosion cracking in service. The objective of the INTERWELD project is to determine the radiation induced material changes that promote stress corrosion cracking in the heat affected zone of austenitic stainless steel welds. To achieve this goal, welds in austenitic stainless steel types AISI 304/347 have been fabricated, respectively. Stress-relief annealing was applied optionally. The pre-characterisation of both the as-welded and stress relieved material conditions comprises the examination of the weld residual stresses by the ring-core-technique and neutron diffraction, the degree of sensitisation by EPR, and the stress corrosion behaviour by SSRT testing in high-temperature water. The weldments will be irratiated to 2 neutron fluence levels and a postirradiation examination will determine micromechanical, microchemical and microstructural changes in the materials. In detail, the evolution of the residual stress levels and the stress corrosion behaviour after irradiation will be determined. Neutron diffraction will be utilized for the first time with respect to neutron irradiated material. In this paper, the current state of the project will be described and discussed. (orig.)

  14. Electrochemical assessment of some titanium and stainless steel impact dental alloys

    International Nuclear Information System (INIS)

    Echavarria, A.; Arroyave, C.

    2003-01-01

    Commercially pure titanium alloy, Ti-6Al-4V alloy and stainless steel screw implants were evaluated in both Ringer and synthetic saliva physiological solutions at body temperature by EIS (Electrochemical Impedance Spectroscopy) with immersion times of 30 d. Results were simulated as a sandwich system composed by four capacitors-resistances connected in series with the solution resistance. A model explaining the results in terms of the porosity and thickness of four different layers, was proposed. (Author) 22 refs

  15. Positron annihilation in hydrogenated and electron-irradiated titanium alloys

    International Nuclear Information System (INIS)

    Mukashev, K.M.; Zaikin, Yu.A.

    2002-01-01

    Important information on hydrogen behavior in titanium can be obtained from studies of radiation damage in previously hydrogenated metal. For this purpose annealed titanium samples were hydrogenated at the temperature 500 deg. C during 1 hour. Then both the original annealed samples and hydrogenated samples were irradiated by 4 MeV electrons in the fluence range 3·10 7 -1·10 19 cm - 2 at the temperature 60 deg. C. It is known that electron irradiation in these conditions predominantly creates vacancy-type defects with an average radius R ν =0.81 Angstrom. It was stated that annihilation probability after electron irradiation of previously hydrogenated titanium samples always has some intermediate values between those characteristic for hydrogenated and irradiated states of previously annealed metal. This is a reason to suppose that radiation defects of the vacancy type in previously hydrogenated titanium combine with hydrogen atoms in favorable conditions of their partial ionization. The estimated value of the average radius for such a complex is R ν =1.1 Angstrom, that is higher than vacancy size but lower than an atom radius. No dose dependence of hydrogen interaction with radiation defects was observed in our experiments.The results of isochrone annealing of the materials under study have shown that the single annealing recovery stage with activation energy E a equal to 1.22 eV is observed in electron irradiated but not previously hydrogenated titanium in the temperature range 170-240 deg. C. Electron irradiation of the previously hydrogenated metal shifts beginning of the first recovery stage to the temperature about 225-230 deg. C and finishes near the temperature 330 deg. C. Therefore, the bound state vacancy-hydrogen in titanium is characterized by higher temperature range of dissociation and annealing with activation energy equal to 1.38 eV. However, subsequent measurements, of the angular distribution of annihilation photons (ADAP) have demonstrated

  16. Effect of ageing on the microstructural stability of cold-worked titanium-modified 15Cr-15Ni-2.5Mo austenitic stainless steel

    International Nuclear Information System (INIS)

    Venkadesan, S.; Bhaduri, A.K.; Rodriguez, P.; Padmanabhan, K.A.

    1992-01-01

    A titanium-modified 15Cr-15Ni-2.5Mo austenitic stainless steel conforming to ASTM A 771 (UNS S 38660), commercially called Alloy D9, is being indigenously developed for application as material for the fuel clad and the hexagonal wrapper for fuel subassemblies of the Prototype Fast Breeder Reactor. As this material would be used in the cold-worked condition and would be subjected to prolonged exposure to elevated service temperatures, the effect of ageing on the microstructural stability was studied as a function of the amount of cold work. The material was given 2.5-30% prior cold work and then aged at temperatures in the range 923 to 1173 K for times ranging from 0.25 to 1000 h. Hardness measurements made before and after ageing were correlated with the Larson-Miller parameter to determine the highest stable prior cold-work level. Optical microscopy was used to study the microstructural changes. The influence of prolonged exposure for two and three years at the operating temperatures of clad and wrapper, on the elevated temperature tensile properties of a 20% prior cold-worked Alloy D9 was also studied through accelerated ageing treatments based on the present parametric approach. (orig.)

  17. Hyper-thermal neutron irradiation field for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1994-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

  18. Full title: Biomechanical comparison between stainless steel, titanium and carbon-fiber reinforced polyetheretherketone volar locking plates for distal radius fractures.

    Science.gov (United States)

    Mugnai, Raffaele; Tarallo, Luigi; Capra, Francesco; Catani, Fabio

    2018-05-25

    As the popularity of volar locked plate fixation for distal radius fractures has increased, so have the number and variety of implants, including variations in plate design, the size and angle of the screws, the locking screw mechanism, and the material of the plates. carbon-fiber reinforced polyetheretherketone (CFR-PEEK) plate features similar biomechanical properties to metallic plates, representing, therefore, an optimal alternative for the treatment of distal radius fractures. three different materials-composed plates were evaluated: stainless steel volar lateral column (Zimmer); titanium DVR (Hand Innovations); CFR-PEEK DiPHOS-RM (Lima Corporate). Six plates for each type were implanted in sawbones and an extra-articular rectangular osteotomy was created. Three plates for each material were tested for load to failure and bending stiffness in axial compression. Moreover, 3 constructs for each plate were evaluated after dynamically loading for 6000 cycles of fatigue. the mean bending stiffness pre-fatigue was significantly higher for the stainless steel plate. The titanium plate yielded the higher load to failure both pre and post fatigue. After cyclic loading, the bending stiffness increased by a mean of 24% for the stainless steel plate; 33% for the titanium; and 17% for the CFR-PEEK plate. The mean load to failure post-fatigue increased by a mean of 10% for the stainless steel and 14% for CFR-PEEK plates, whereas it decreased (-16%) for the titanium plate. Statistical analysis between groups reported significant values (p plastic deformation, and lower load to failure. N/A. Copyright © 2018. Published by Elsevier Masson SAS.

  19. Applications of neutron irradiation

    International Nuclear Information System (INIS)

    Ito, Yasuo

    1999-01-01

    The present state of art of applications of neutron irradiation is overviewed taking neutron activation analysis, prompt gamma-ray analysis, fission/alpha track methods, boron neutron capture therapy as examples. What is common among them is that the technologies are nearly matured for wide use by non- nuclear scientists. But the environment around research reactors is not prospective. These applications should be encouraged by incorporating in the neutron science society. (author)

  20. Modified fuel channel for sample irradiation at the RB reactor

    International Nuclear Information System (INIS)

    Pesic, M.; Markovic, H.; Sokcic, M.; Miric, I.; Prokic, M.; Strugar, P.

    1983-01-01

    Fuel channel of 80% enriched UO 2 at RB reactor in Boris Kidric Institute of nuclear sciences is modified for sample irradiation in the fast neutron field. Maximum sample diameter is 25 mm and length up to 100 mm. Characteristics of neutron as well as gamma radiation fields of this new experimental channel are investigated. In the center of channel, the main contribution to the total neutron absorbed dose i.e. 0.29 Gy per 1 Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to epithermal and thermal neutrons respectively. At the same time gamma absorption dose is 0.35 Gy. The development of experimental fuel channel GRK has wide possibility for utilization, from electronic components fast neutron studies, dosimeters testing, to cross section measurements for fast neutron energies. (author)

  1. Precipitation response of austenitic stainless steel to simulated fusion irradiation

    International Nuclear Information System (INIS)

    Maziasz, P.J.

    1978-01-01

    The precipitation response of annealed type 316 stainless steel irradiated in HFIR is studied and compared to previously observed thermal aging and fast reactor irradiation responses. Irradiation in HFIR simultaneously produces high levels of helium and displacement damage and partially simulates a fusion environment. Samples have been irradiated at temperatures from 550 to 680 0 C to fluences producing up to 3300 appm He and 47 dpa

  2. The effect of molten salt on high temperature behavior of stainless steel and titanium alloy with the presence of water vapor

    Science.gov (United States)

    Baharum, Azila; Othman, Norinsan Kamil; Salleh, Emee Marina

    2018-04-01

    The high temperature oxidation experiment was conducted to study the behavior of titanium alloy Ti6A14V and stainless steel 316 in Na2SO4-50%NaCl + Ar-20%O2 (molten salt) and Na2SO4-50%NaCl + Ar-20%O2 + 12% H2O (molten salt + water vapor) environment at 900°C for 30 hours using horizontal tube furnace. The sample then was investigated using weight change measurement analysis and X-ray diffraction (XRD) analysis to study the weight gained and the phase oxidation that occurred. The weight gained of the titanium alloy was higher in molten salt environment compared to stainless steel due to the rapid growth in the oxide scale but showed almost no change of weight gained upon addition of water vapor. This is due to the alloy was fully oxidized. Stainless steel showed more protection and better effect in molten salt environment compared to mixed environment showed by slower weight gain and lower oxidation rate. Meanwhile, the phase oxidation test of the samples showed that the titanium alloy consist of multi oxide layer of rutile (TiO2) and Al2O3 on the surface of the exposed sample. While stainless steel show the formation of both protective Cr-rich oxide and non-protective Fe-rich oxide layer. This can be concluded that stainless steel is better compared to Ti alloy due to slow growing of chromia oxide. Therefore it is proven that stainless steel has better self-protection upon high temperature exposure.

  3. Characterization of diffusion bonded joint between titanium and 304 stainless steel using a Ni interlayer

    International Nuclear Information System (INIS)

    Kundu, S.; Chatterjee, S.

    2008-01-01

    Solid-state diffusion bonded joints were prepared between commercially pure titanium and 304 stainless steel with nickel as an intermediate material in the temperature range of 800-950 deg. C for 10.8 ks under a 3 MPa uniaxial pressure in vacuum. The interface microstructures and reaction products of the transition joints were investigated by optical and scanning electron microscopy. Up to 850 deg. C processing temperature, a 300-μm nickel interlayer completely restricts the diffusion of titanium to stainless steel. However, the nickel interlayer cannot block the diffusion of Ti to the stainless side and λ + χ + α-Fe, λ + FeTi and λ + FeTi + β-Ti phase mixtures are formed at the SS-Ni interface, when bonding was processed at 900 deg. C and above. These reaction products were confirmed by X-ray diffraction. A maximum tensile strength of ∼ 270 MPa and shear strength of ∼ 194 MPa, along with 6.2% ductility, were obtained for the diffusion bonded joint processed at 850 deg. C. Fracture surface observation in SEM using EDS demonstrates that failure occurred through the Ni-Ti interface of the joints when processed up to 850 deg. C and through the SS-Ni interface when processed at and above 900 deg. C

  4. X-ray photoelectron spectroscopy studies of hard coatings formed by titanium on 304 stainless steel

    International Nuclear Information System (INIS)

    Nair, M.R.; Kothari, D.C.; Rangwala, A.A.; Lal, K.B.; Prabhawalkar, P.D.; Raole, P.M.

    1986-01-01

    Titanium ions are implanted (at 30 keV) in 304 stainless steel to a dose of 1.8x10 17 ions cm -2 using 15 μA cm -2 and 5 μA cm -2 beam current densities for specimens 2 and 3 respectively. X-ray photoelectron spectroscopy (XPS) measurements are performed at different temperatures. The microhardness of implanted and unimplanted specimens is also measured. In specimen 2 the microhardness does not increase significantly and XPS measurements give evidence of carburized surface alloy formation. At 250 0 C TiO 2 is detected on the surface and it migrates into the bulk phase above 350 0 C. In specimen 3 the XPS measurements exhibit an absence of iron owing to the radiation-induced segregation of titanium on the surface. This specimen shows an increase in microhardness. The XPS measurements reveal a layer of (TiC x -C) on the surface which is suggested to be responsible for the increase in microhardness. Upon heating, TiC x is seen to move into the bulk phase and the carbon concentration is increased. These changes occurring at higher temperatures are suggested as having an effect on the wear-resistant properties of titanium-implanted 304 stainless steel. (orig.)

  5. Analysis of SCC initiation/propagation behavior of stainless steels in LWR environments

    International Nuclear Information System (INIS)

    Saito, Koichi; Kuniya, Jiro

    1999-01-01

    This paper presents a method to analyze initiation and propagation behavior of stress corrosion cracking (SCC) of stainless steels on the basis of a new prediction algorithm in which the initiation period and propagation period of SCC under irradiation conditions are considered from a practical viewpoint. The prediction algorithm is based on three ideas: (1) threshold neutron fluence of radiation-enhanced SCC (RESCC), (2) equivalent critical crack depth, and (3) threshold stress intensity factor for SCC (K ISCC ). SCC initiation/propagation behavior in light water reactor (LWR) environments is analyzed by incorporating model equations on irradiation hardening, irradiation-enhanced electrochemical potentiokinetic reactivation (EPR) and irradiation stress relaxation that are phenomena peculiar to neutron irradiation. The analytical method is applied to predict crack growth behavior of a semi-elliptical surface crack in a flat plane that has an arbitrary residual stress profile; specimens are sensitized type 304 stainless steels which had been subjected to neutron irradiation in high temperature water. SCC growth behavior of a semi-elliptical surface crack was greatly dependent on the distribution of residual stress in a flat plane. When residual stress at the surface of the flat plane was relatively small, the method predicted SCC propagation did not take place. (author)

  6. Evaluation of Antibacterial Activity of Titanium Surface Modified by PVD/PACVD Process.

    Science.gov (United States)

    Ji, Min-Kyung; Lee, Min-Joo; Park, Sang-Won; Lee, Kwangmin; Yun, Kwi-Dug; Kim, Hyun-Seung; Oh, Gye-Jeong; Kim, Ji-Hyun; Lim, Hyun-Pil

    2016-02-01

    The aim of this study was to evaluate the response of Streptococcus mutans (S. mutans) via crystal violet staining assay on titanium surface modified by physical vapor deposition/plasma assisted chemical vapor deposition process. Specimens were divided into the following three groups: polished titanium (control group), titanium modified by DC magnetron sputtering (group TiN-Ti), and titanium modified by plasma nitriding (group N-Ti). Surface characteristics of specimens were observed by using nanosurface 3D optical profiler and field emission scanning electron microscope. Group TiN-Ti showed TiN layer of 1.2 microm in thickness. Group N-Ti was identified as plasma nitriding with X-ray photoelectron spectroscopy. Roughness average (Ra) of all specimens had values 0.05). Within the process condition of this study, modified titanium surfaces by DC magnetron sputtering and plasma nitriding did not influence the adhesion of S. mutans.

  7. Alterations in water and electrolyte absorption in the rat colon following neutron irradiation: influence of neutron component and irradiation dose.

    Science.gov (United States)

    Dublineau, I; Ksas, B; Joubert, C; Aigueperse, J; Gourmelon, P; Griffiths, N M

    2002-12-01

    To study the absorptive function of rat colon following whole-body exposure to neutron irradiation, either to the same total dose with varying proportion of neutrons or to the same neutron proportion with an increasing irradiation dose. Different proportions of neutron irradiation were produced from the reactor SILENE using a fissile solution of uranium nitrate (8, 47 and 87% neutron). Water and electrolyte fluxes were measured in the rat in vivo under anaesthesia by insertion into the descending colon of an agarose gel cylinder simulating the faeces. Functional studies were completed by histological analyses. In the first set of experiments, rats received 3.8 Gy with various neutron percentages and were studied from 1 to 14 days after exposure. In the second set of experiments, rats were exposed to increasing doses of irradiation (1-4Gy) with a high neutron percentage (87%n) and were studied at 4 days after exposure. The absorptive capacity of rat colon was diminished by irradiation at 3-5 days, with a nadir at 4 days. The results demonstrate that an increase in the neutron proportion is associated with an amplification of the effects. Furthermore, a delay in the re-establishment of normal absorption was observed with the high neutron proportion (87%n). A dose-dependent reduction of water absorption by rat colon was also observed following neutron irradiation (87%n), with a 50% reduction at 3 Gy. Comparison of this dose-effect curve with the curve obtained following gamma (60)Co-irradiation indicates an RBE of 2.2 for absorptive colonic function in rat calculated at 4 days after exposure.

  8. Study of neutron irradiation on F82H alloys by Mössbauer spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Huang, S.S., E-mail: h.shaosong@ht8.ecs.kyoto-u.ac.jp; Kitao, S.; Kobayashi, Y.; Yoshiie, T.; Xu, Q.; Sato, K.; Seto, M.

    2015-01-15

    The effects of neutron irradiation on F82H ferritic/martensitic stainless steel and its model alloys were studied by Mössbauer spectroscopy. The microstructural damage mechanisms of these alloys, during the void incubation period were interpreted using the short range order (SRO) parameters. Results show that within Fe–8Cr alloy, the atoms in the nearest neighbor (NN) of the Fe nuclei were inhomogeneous, prior to irradiation. A configuration trapping model of Cr supported the negative average SRO observed for the NN shells in our Fe–Cr alloys. We found that irradiation also accelerated the SRO in Fe–8Cr through a diffusion mechanism, where Cr atom repulsion was concentration dependent. Finally, comparative studies were conducted on F82H model alloys using the present Mössbauer measurements and our previously reported work on positron annihilation spectroscopy, which further established that irradiation of the alloys promoted the growth of a M{sub 23}C{sub 6} complex.

  9. Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

    International Nuclear Information System (INIS)

    Liu, H.B.

    1996-01-01

    The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 10 10 n/cm 2 ·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10 -11 cGy·cm 2 /n th ) and fast neutron (0.58 x 10 -11 cGy·cm 2 /n th ) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed

  10. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  11. Effects of solute interstitial elements on swelling of stainless steel

    International Nuclear Information System (INIS)

    Stiegler, J.O.; Leitnaker, J.M.; Bloom, E.E.

    1975-01-01

    High-purity stainless steel (HPS), equivalent to type 316 stainless steel in major alloy elements but with greatly reduced interstitial elements and manganese contents, was irradiated in the temperature range 725 to 875 K to fluences ranging from 1.0 to 3.5 x 10 26 neutrons/m 2 (>0.1 MeV). The HPS swelled 20 to 50 times more than commercial grade 316 stainless steel (316 SS), and about the same as commercial-purity nickel, which has about the same interstitial content as HPS. A fine-grained 316 SS in which interstitial elements but not manganese were precipitated by thermomechanical treatments also showed exaggerated swelling, approaching that of HPS, which suggests that swelling in commercial stainless steels is retarded by small amounts of interstitial elements normally present in them and not by the major alloying elements. Interstitials tend to precipitate from solution during irradiation, and bulk extractions of precipitate particles were made to evaluate the extent of the precipitation reactions. At both 643 and 853 K precipitation was clearly enhanced by irradiation significantly enough to alter the matrix composition, which suggests that swelling may be increased at high fluences over that predicted by extrapolation of lower fluence data. These observations are discussed in terms of potential behaviour of fuel cladding materials and of the validity and interpretation of accelerated schemes for simulating neutron damage. (author)

  12. Neutron irradiation control in the neutron transmutation doping process in HANARO using SPND

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Gi-Doo; Kim, Myong-Seop [Korea Atomic Energy Research Institute, Yuseong, Daejeon, 305-353, (Korea, Republic of)

    2015-07-01

    The neutron irradiation control method by using self-powered neutron detector (SPND) is developed for the neutron transmutation doping (NTD) application in HANARO. An SPND is installed at a fixed position of the upper part of the sleeve in HANARO NTD hole for real-time monitoring of the neutron irradiation. It is confirmed that the SPND is significantly affected by the in-core condition and surroundings of the facility. Furthermore, the SPND signal changes about 15% throughout a whole cycle according to the change of the control rod position. But, it is also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation can is not so big while moving of the control rod. Accordingly, the relationship between the ratio of the neutron flux to the SPND signal output and the control rod position is established. In this procedure, the neutron flux measurement by using zirconium foil is utilized. The real NTD irradiation experiments are performed using the established relationship. The irradiated neutron fluence can be controlled within ±1.3% of the target one. The mean value of the irradiation/target ratio of the fluence is 0.9992, and the standard deviation is 0.0071. Thus, it is confirmed that the extremely accurate irradiation would be accomplished. This procedure can be useful for the SPND application installed at the fixed position to the field requiring the extremely high accuracy. (authors)

  13. Conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at SINQ

    International Nuclear Information System (INIS)

    Zanini, L.; Baluc, N.; Simone, A. De; Eichler, R.; Joray, S.; Manfrin, E.; Pouchon, M.; Rabaioli, S.; Schumann, D.; Welte, J.; Zhernosekov, K.

    2011-12-01

    This comprehensive, illustrated report by the Paul Scherrer Institute PSI in Switzerland documents the proposals concerning the conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at the PSI's Swiss Spallation Neutron Source SINQ facility. The need for fast neutron irradiation is discussed and the possibility of using SINQ as a fast neutron irradiation facility is considered. The production of isotopes, tracers and medical isotopes is discussed, as are fission and fusion reactor technologies. The characteristics of the neutron spectrum in SINQ are discussed. The neutronic and radioprotection calculations for an irradiation station at SINQ are looked at in detail and extensive examples of work done and results obtained are presented and discussed. Radioprotection issues are also looked at. Further contributions in the report cover the hot/cold irradiation station in the SINQ target. An appendix provides detailed drawings of the facility's pneumatic delivery system

  14. Microstructural changes of Y-doped V-4Cr-4Ti alloys after ion and neutron irradiation

    Directory of Open Access Journals (Sweden)

    H. Watanabe

    2016-12-01

    Full Text Available High-purity Y-doped V-4Cr-4Ti alloys (0.1–0.2wt. % Y, manufactured by the National Institute for Fusion Science (NIFS, were used for this study. Heavy-ion and fission-neutron irradiation was carried out at temperatures 673–873K. During the ion irradiation at 873K, the microstructure was controlled by the formation of Ti(C,O,N precipitates lying on the (100 plane. Y addition effectively suppressed the growth of Ti(C,O,N precipitates, especially at lower dose irradiation to up to 4 dpa. However, at higher dose levels (12.0 dpa, the number density was almost at the same levels irrespective of the presence of Y. After neutron irradiation at 873K, fine titanium oxides were also observed in all V alloys. However, smaller oxide sizes were observed in the Y-doped samples under the same irradiation conditions. The detailed analysis of EDS showed that the center of the Ti(C,O,N precipitates was mainly enriched by nitrogen. The results showed that the contribution of not only oxygen atoms picked up from the irradiation environment but also nitrogen atoms is essential to understand the microstructural evolution of V-4Cr-4Ti-Y alloys.

  15. A Modified Porous Titanium Sheet Prepared by Plasma-Activated Sintering for Biomedical Applications

    Directory of Open Access Journals (Sweden)

    Yukimichi Tamaki

    2010-01-01

    Full Text Available This study aimed to develop a contamination-free porous titanium scaffold by a plasma-activated sintering within an originally developed TiN-coated graphite mold. The surface of porous titanium sheet with or without a coated graphite mold was characterized. The cell adhesion property of porous titanium sheet was also evaluated in this study. The peak of TiC was detected on the titanium sheet processed with the graphite mold without a TiN coating. Since the titanium fiber elements were directly in contact with the carbon graphite mold during processing, surface contamination was unavoidable event in this condition. The TiC peak was not detectable on the titanium sheet processed within the TiN-coated carbon graphite mold. This modified plasma-activated sintering with the TiN-coated graphite mold would be useful to fabricate a contamination-free titanium sheet. The number of adherent cells on the modified titanium sheet was greater than that of the bare titanium plate. Stress fiber formation and the extension of the cells were observed on the titanium sheets. This modified titanium sheet is expected to be a new tissue engineering material in orthopedic bone repair.

  16. Microstructural evolution in modified 9Cr-1Mo ferritic/martensitic steel irradiated with mixed high-energy proton and neutron spectra at low temperatures

    International Nuclear Information System (INIS)

    Sencer, B.H.; Garner, F.A.; Gelles, D.S.; Bond, G.M.; Maloy, S.A.

    2002-01-01

    Modified 9Cr-1Mo ferritic/martensitic steel was exposed at 32-57 deg. C to a mixed proton/neutron particle flux and spectrum at the Los Alamos Neutron Science Center. The microstructure of unirradiated 9Cr-1Mo consists of laths, dislocations and carbides. Examination of electron diffraction patterns obtained from extraction replicas of unirradiated 9Cr-1Mo revealed that the precipitate microstructure was primarily dominated by M 23 C 6 carbides. The post-irradiation microstructure contained black-spot damage in addition to precipitates and dislocations. Examination of electron diffraction patterns revealed diffuse rings from M 23 C 6 carbides, indicating amorphization and/or nanocrystallinity. Crystalline MC carbides were also found. No cavity formation was found although a significant amount of helium and hydrogen generation had been generated. TEM-EDS examination of extraction replicas for carbides from unirradiated and irradiated samples did not show any detectable changes in composition of either M 23 C 6 or MC carbides. There was also no evident change in carbide size. Lattice images of M 23 C 6 carbides revealed an amorphous microstructure following irradiation, but MC carbides were still crystalline

  17. Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) of austenitic stainless steels in oxygenated high temperature water was studied. The IASCC failure has been considered as a degradation phenomenon potential not only in the present light water reactors but rather common in systems where the materials are exposed simultaneously to radiation and water environments. In this study, effects of the material and environmental factors on the IASCC of austenitic stainless steels were investigated in order to understand the underlying mechanism. The following three types of materials were examined: a series of model alloys irradiated at normal water-cooled research reactors (JRR-3M and JMTR), the material irradiated at a spectrally tailored mixed-spectrum research reactor (ORR), and the material sampled from a duct tube of a fuel assembly used in the experimental LMFBR (JOYO). Post-irradiation stress corrosion cracking tests in a high-temperature water, electrochemical corrosion tests, etc., were performed at hot laboratories. Based on the results obtained, analyses were made on the effects of alloying/impurity elements, irradiation/testing temperatures and material processing, (i.e., post-irradiation annealing and cold working) on the cracking behavior. On the basis of the analyses, possible remedies against IASCC in the core internals were discussed from viewpoints of complex combined effects among materials, environment and processing factors. (author). 156 refs.

  18. The effects of Ti implantation on corrosion and adhesion of TiN coated stainless steel

    Science.gov (United States)

    Baba, K.; Nagata, S.; Hatada, R.; Daikoku, T.; Hasaka, M.

    1993-06-01

    Thin titanium nitride (TiN) films of 40 and 70 nm in thickness were deposited on austenitic-type 304 stainless steel substrates by a rf ion plating process, and these specimens were irradiated with 70 kV titanium ions at a fluence of 1 × 10 17/cm 2 by use of MEVVA IV metallic ion source at room temperature. After that TiN films of 2 μm were deposited by the same method. The results of X-ray photoelectron spectroscopy and Auger electron spectroscopy revealed that implanted titanium penetrated into the substrate and interfacial mixing was verified. The adhesion strength was estimated by a scratch test. It was found that ion implantation can enhance the adhesion strength between the film and the substrate. The corrosion resistance of the specimens was evaluated in aqueous solutions of sulfuric acid by an electrochemical method. Titanium implantation was extremely effective in suppressing the anodic dissolution of stainless steel.

  19. Effects of irradiation on the fracture behavior of austenitic stainless steels

    International Nuclear Information System (INIS)

    Grossbeck, M.L.; Stiegler, J.O.; Holmes, J.J.

    1977-01-01

    Fracture in irradiated materials occurs by mechanisms which occur in unirradiated materials in addition to mechanisms related to irradiation phenomena. The paper examines radiation effects in austenitic stainless steels for use as core structural materials in fast breeder reactors

  20. Neutron irradiation of bacteriophage λ

    International Nuclear Information System (INIS)

    Bozin, D.; Milosevic, M. . E-mail address of corresponding author: bozinde@vin.bg.ac.yu

    2005-01-01

    Double strand breaks (DSB) are the most dangerous lesions in DNA caused by irradiation, but many other lesions, usually called mutations, have not been clearly identified. These lesions, like DSB, can be the source of serious chromosomal damages and finally - cell death. Growing interest in heavy particles for radiotherapy and radioprotection encourages the search of the molecular basis of their action. In this respect, we chose bacteriophage λ1390 as the model system for the study of consequences of neutron irradiation. This derivative of λ phage possesses an unique ability to reversibly reorganize their genome in response to various selective pressures. The phages were irradiated with 13 Gy of mixed neutrons (7.5 Gy from fast and 5.6 Gy from thermal neutrons) and phages genomes were tested to DSB and mutations. Additionally, the stability of λ capsid proteins were tested. After all tests, we can conclude that, under our conditions, low flux of neutrons does not induce neither DNA strand break or DNA mutation nor the stability of λ capsid proteins. (author)

  1. Grain-boundary microchemistry and intergranular cracking of irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Chung, H.M.; Ruther, W.E.; Sanecki, J.E.; Kassner, T.F.

    1993-01-01

    Constant-extension-rate tensile tests and grain-boundary analysis by Auger electron spectroscopy were conducted on high and commercial-purity (HP and CP) Type 304 stainless steel (SS) specimens from irradiated boiling-water reactor (BWR) components to identify the mechanisms of irradiation-assisted stress corrosion cracking (IASCC). Contrary to previous beliefs, susceptibility to intergranular fracture could not be correlated with radiation-induced segregation of impurities such as Si, P, C, or S, but a correlation was obtained with grain-boundary Cr concentration, indicating a role for Cr depletion. Detailed analysis of grain-boundary chemistry was conducted on BWR neutron absorber tubes that were fabricated from two similar heats of HP Type 304 SS of virtually identical bulk chemical composition but exhibiting a significant difference in susceptibility to IASCC after irradiation to ∼2 x 10 21 n/cm 2 (E > 1 MeV). Grain-boundary concentrations of Cr Ni, Si, P, S, and C of the cracking-resistant and -susceptible HP heats were virtually identical. However, grain boundaries of the cracking-resistant material contained less N and more B and Li than those of the cracking-susceptible material. This observation indicates that, besides the deleterious effect of grain-boundary Cr depletion, a synergism between grain-boundary segregation of N and B and transmutation to H and Li plays an important role in IASCC

  2. Evaluation of non-conformities of hip prostheses made of titanium alloys and stainless steel; Avaliacao de nao conformidades de proteses de quadril fabricadas com ligas de titanio e aco inox

    Energy Technology Data Exchange (ETDEWEB)

    Bezerra, Ewerton de Oliveira Teotonio; Nascimento, Jose Jeferson da Silva; Luna, Carlos Bruno Barreto; Morais, Crislene Rodrigues da Silva; Campos, Karla Valeria Miranda de, E-mail: ewerton.teotonio@hotmail.com, E-mail: brunobarretodemaufcg@hotmail.com [Universidade Federal de Campina Grande (UAEMa/CCT/UFCG), PB (Brazil). Unidade Academinca de Engenharia de Materiais

    2017-01-15

    A large number of metallic alloys has satisfactory behavior when used to manufacture implants for hip prostheses. However, they must be in conformity with standards, to ensure their quality for long periods without losing its functionality. Therefore, this paper aims to study the non-conformities in two hip prostheses, one of titanium and other stainless steel according to standards. The implants studied passed by X-ray diffraction (XRD), X-ray fluorescence, tensile test and optical microscopy (OM). Specimens for the tensile test were made according to ASTM E 8M, as well, MO samples passed by metallographic procedure. The results evidenced that some chemical compositions showed in relation to the standards. The XRD analysis showed peaks of austenite and absence of ferrite for the stainless steel, while the titanium alloy presents an alpha phase (HCP) more significant than the beta phase (BCC). The stainless steel alloys and titanium have yield strength and tensile strength that meet the standards. On the other hand, the elastic modulus of the titanium alloy and stainless steel, comes to be ten times greater than the human bone. Therefore, the high modulus of elasticity of the alloys, favors bone resorption problems. The stainless steel microstructure is typical of an austenitic matrix, while the titanium alloy presents α + β microstructure. (author)

  3. Temper-bead repair-welding of neutron-irradiated reactor (pressure) vessel by low-heat-input TIG and YAG laser welding

    International Nuclear Information System (INIS)

    Nakata, Kiyotomo; Ozawa, Masayoshi; Kamo, Kazuhiko

    2006-01-01

    Weldability in neutron-irradiated low alloy steel for reactor (pressure) vessel has been studied by temper-bead repair-welding of low-heat-input TIG and YAG laser welding. A low alloy steel and its weld, and stainless steel clad and nickel (Ni)-based alloy clad were irradiated in a materials test reactor (LVR-15, Czech Republic) up to 1.4 x 10 24 n/m 2 (>1 MeV) at 290degC, which approximately corresponds to the maximum neutron fluence of 60-year-operation plants' vessels. The He concentration in the irradiated specimens was estimated to be up to 12.9 appm. The repair-welding was carried out by TIG and YAG laser welding at a heat input from 0.06 to 0.86 MJ/m. The mechanical tests of tensile, impact, side bend and hardness were carried out after the repair-welding. Cracks were not observed in the irradiated low alloy steel and its weld by temper-bead repair-welding. Small porosities were formed in the first and second layers of the repair-welds of low alloy steel (base metal). However, only a few porosities were found in the repair-welds of the weld of low alloy steel. From the results of mechanical tests, the repair-welding could be done in the irradiated weld of low alloy steel containing a He concentration up to 12.9 appm, although repair-welding could be done in base metal of low alloy steel containing up to only 1.7 appmHe. On the other hand, cracks occurred in the heat affected zones of stainless steel and Ni-based alloy clads by repair-welding, except by YAG laser repair-welding at a heat input of 0.06 MJ/m in stainless steel clad containing 1.7 appmHe. Based on these results, the determination processes were proposed for optimum parameters of repair-welding of low alloy steel and clad used for reactor (pressure) vessel. (author)

  4. Effects of residual stress on irradiation hardening in stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, N.; Kondo, K.; Kaji, Y. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Miwa, Y. [Nuclear Energy and Science Directorate, Japan Atomic Energy Agency, Tokai-mura, Ibaraki-ken (Japan)

    2007-07-01

    Full text of publication follows: Structural materials in fusion reactor with water cooling system will undergo corrosion in aqueous environment and heavier irradiation than that in LWR. Irradiation assisted stress corrosion (IASCC) may be induced in stainless steels exposed in these environment for a long term of reactor operation. The IASCC is considered to be caused in a welding zone. It is difficult to predict and estimate the IASCC, because several irradiation effects (irradiation hardening, swelling, irradiation induced stress relaxation, etc) work intricately. Firstly, effects of residual stress on irradiation hardening were investigated in stainless steels. Specimens used in this study were SUS316 and SUS316L. By bending deformation, the specimens with several % plastic strain, which corresponds to weld residual stress, were prepared. Ion irradiations of 12 MeV Ni{sup 3+} were performed at 330, 400 and 550 deg. C to 45 dpa in TIARA facility at JAEA. No bent specimen was simultaneously irradiated with the bent specimen. The residual stress was estimated by X-ray residual stress measurements before and after the irradiation. The micro-hardness was measured by using nano-indenter. The irradiation hardening and the stress relaxation were changed by irradiation under bending deformation. The residual stress did not relax even for the case of the higher temperature aging at 500 deg. C for the same time of irradiation. The residual stress after ion irradiation, however, relaxed at these experimental temperatures in SUS316L. The hardness was obviously suppressed in bent SUS316L irradiated at 300 deg. C to 6 or 12 dpa. It was evident that irradiation induced stress relaxation occasionally suppressed the irradiation hardening in SUS316L. (authors)

  5. Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

    International Nuclear Information System (INIS)

    Yabuuchi, Kiyohiro; Kuribayashi, Yutaka; Nogami, Shuhei; Kasada, Ryuta; Hasegawa, Akira

    2014-01-01

    Ion irradiation experiments are useful for investigating irradiation damage. However, estimating the irradiation hardening of ion-irradiated materials is challenging because of the shallow damage induced region. Therefore, the purpose of this study is to prove usefulness of nanoindentation technique for estimation of irradiation hardening for ion-irradiated materials. SUS316L austenitic stainless steel was used and it was irradiated by 1 MeV H + ions to a nominal displacement damage of 0.1, 0.3, 1, and 8 dpa at 573 K. The irradiation hardness of the irradiated specimens were measured and analyzed by Nix–Gao model. The indentation size effect was observed in both unirradiated and irradiated specimens. The hardness of the irradiated specimens changed significantly at certain indentation depths. The depth at which the hardness varied indicated that the region deformed by the indenter had reached the boundary between the irradiated and unirradiated regions. The hardness of the irradiated region was proportional to the inverse of the indentation depth in the Nix–Gao plot. The bulk hardness of the irradiated region, H 0 , estimated by the Nix–Gao plot and Vickers hardness were found to be related to each other, and the relationship could be described by the equation, HV = 0.76H 0 . Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials

  6. Deformation and Fracture Properties in Neutron Irradiated Pure Mo and Mo Alloys

    International Nuclear Information System (INIS)

    Byun, T.S.; Snead, L.; Li, M.; Cockeram, B.V.

    2007-01-01

    Full text of publication follows: The evolution in microstructural and mechanical properties was investigated for molybdenum and molybdenum alloys after high temperature neutron irradiation. Test materials include oxide dispersion-strengthened (ODS) molybdenum alloy, molybdenum- 0.5% titanium-0.1% zirconium (TZM) alloy, and low carbon arc-cast (LCAC) molybdenum. Tensile specimens were irradiated in high flux isotope reactor (HFIR) at temperatures in the range ∼300 - 1000 deg. C to neutron fluences of 2.28 - 24.7 x 10 25 n/m 2 (E>0.1 MeV) or 1.2-13.1 dpa. Tensile tests were performed at temperatures ranging from -150 deg. C to 1000 deg. C. To evaluate irradiation effects, true stress parameters (yield stress, plastic instability stress, and true fracture stress) and ductility parameters (uniform strain, fracture strain, and reduction area) were compared for both irradiated and non-irradiated materials. Fracture toughness was also evaluated from the fracture stress and fracture strain data using a fracture strain model. The fracture strain was used to determine the ductile-to-brittle transition temperature (DBTT). Results indicate that irradiation in the temperature range of 600 - 800 deg. C hardened the materials by up to 70%, while the irradiation hardening outside this temperature range was much lower (<40%). The plastic instability stress was strongly dependent on test temperature; however, it was nearly independent of irradiation dose and temperature. It was also found that the true fracture stress was dependent on test temperature. The true fracture stress was not significantly influenced by irradiation at elevated and high test temperatures; however, it was decreased significantly at sub-zero temperatures after irradiation due to material embrittlement. The DBTT for 600 deg. C irradiated ODS molybdenum alloy was found to be about room temperature or lower, and among the test materials the ODS alloy showed the highest resistance to irradiation embrittlement

  7. Mechanisms of stress relief cracking in titanium stabilised austenitic stainless steel

    International Nuclear Information System (INIS)

    Chabaud-Reytier, M.; Allais, L.; Caes, C.; Dubuisson, P.; Pineau, A.

    2003-01-01

    The heat affected zone (HAZ) of AISI 321 welds may exhibit a serious form of cracking during service at high temperature. This form of damage, called 'stress relief cracking', is known to be due to work hardening but also to aging due to Ti(C,N) precipitation on dislocations which modifies the mechanical behaviour of the HAZ. The present study aims to analyse the latter embrittlement mechanism in one specific heat of 321 stainless steel. To this end, different HAZs are simulated using an annealing heat-treatment, followed by various cold rolling and aging conditions. Then, we study the effects of work hardening and aging on Ti(C,N) precipitation, on the mechanical (hardness, tensile and creep) behaviour of the simulated HAZs and on their sensitivity to intergranular crack propagation through stress relaxation tests performed on pre-cracked CT type specimens tested at 600 deg. C. It is shown that work hardening is the main parameter of the involved mechanism but that aging does not promote crack initiation although it leads to titanium carbide precipitation. Therefore, the role of Ti(C,N) precipitation on stress relief cracking mechanisms is discussed. An attempt is made to show that solute drag effects are mainly responsible for this form of intergranular damage, rather than Ti(C,N) precipitation

  8. Design of an irradiation facility with thermal, epithermal and fast neutron beams

    International Nuclear Information System (INIS)

    Pfister, G.; Bernnat, W.; Seidel, R.; Schatz, A.K.; Wagner, F.M.; Waschkowski, W.; Schraube, H.

    1992-01-01

    The main features of a neutron irradiation facility to be installed at the planned research reactor FRM-II are presented. In addition to the operational possibilities of the existing facility at the reactor FRM-I, the new facility will produce quasi-monoenergetic neutron fields and a neutron beam in the keV region whose spectrum can be modified by application of suitable filters and scatterers. For this beam, which is well suited for boron capture therapy, calculated boron reaction rates inside a phantom and an experimental verification of the calculations at the existing facility are presented. (orig.) [de

  9. Neutron Dosimetry and Irradiation of Solids; Dosimetrie des neutrons et irradiation des solides

    Energy Technology Data Exchange (ETDEWEB)

    Perriot, G; Schmitt, A P [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1962-07-01

    Results of work at C.E.A. from 1958 to 1960 are reviewed. The possibilities offered by classical dosimetry methods are discussed. The tests which led to the utilization, for fast neutron dosimetry, of resistivity variations induced in solid W by such neutrons are described. Experimental W irradiation results led to a definition of neutron efficiency which describes the relations between neutron energy and their effects on materials. Possibilities offered by detectors which make use of radiation damage and are sensitive to neutrons at keV energies were explored. In other work, the principal French reactors were classified according to their ability to produce damage in materials such as W. (authors) [French] Dans ce rapport on a presente les resultats essentiels de travaux qui ont ete effectues de 1958 a 1980 par des chercheurs du CEA issus de differents services. En meme temps qu'une revue des possibilites offertes a l'epoque par les methodes classiques de dosimetrie (utilisation des detecteurs par activation), on a decrit les essais qui devaient permettre d'utiliser, a la dosimetrie les neutrons rapides, les variations de resistivite qu'ils creent dans un corps solide (tungstene). L'irradiation du tungstene a montre l'importance qu'il y avait a definir 'l'efficacite' des neutrons, c'est-a-dire leur aptitude plus ou moins grande, selon leur energie, a creer des defauts dans les materiaux. L'efficacite d'un emplacement d'irradiation se trouvant liee au spectre neutronique, on a vu les difficultes qu'il y avait a utiliser les detecteurs par activation des qu'on n'avait plus affaire a un spectre en 1/E ou de fission et on a pu entrevoir les possibilites offertes par les detecteurs utilisant la creation des defauts qui repondent a tous les neutrons d'energies, superieures a quelques keV. Enfin, on a classe les principaux types de Piles Francaises selon leur aptitude a creer plus ou moins rapidement des dommages dans des materiaux comme le tungstene. (auteur)

  10. Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

    Science.gov (United States)

    Kim, Myong-Seop; Park, Sang-Jun

    2009-08-01

    Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

  11. Chemical reactions induced by fast neutron irradiation

    International Nuclear Information System (INIS)

    Katsumura, Y.

    1989-01-01

    Here, several studies on fast neutron irradiation effects carried out at the reactor 'YAYOI' are presented. Some indicate a significant difference in the effect from those by γ-ray irradiation but others do not, and the difference changes from subject to subject which we observed. In general, chemical reactions induced by fast neutron irradiation expand in space and time, and there are many aspects. In the time region just after the deposition of neutron energy in the system, intermediates are formed densely and locally reflecting high LET of fast neutrons and, with time, successive reactions proceed parallel to dissipation of localized energy and to diffusion of the intermediates. Finally the reactions are completed in longer time region. If we pick up the effects which reserve the locality of the initial processes, a significant different effect between in fast neutron radiolysis and in γ-ray radiolysis would be derived. If we observe the products generated after dissipation and diffusion in longer time region, a clear difference would not be observed. Therefore, in order to understand the fast neutron irradiation effects, it is necessary to know the fundamental processes of the reactions induced by radiations. (author)

  12. Hair dosimetry following neutron irradiation.

    Science.gov (United States)

    Lebaron-Jacobs, L; Gaillard-Lecanu, E; Briot, F; Distinguin, S; Boisson, P; Exmelin, L; Racine, Y; Berard, P; Flüry-Herard, A; Miele, A; Fottorino, R

    2007-05-01

    Use of hair as a biological dosimeter of neutron exposure was proposed a few years ago. To date, the (32)S(n,p)(32)P reaction in hair with a threshold of 2.5 MeV is the best choice to determine the fast neutron dose using body activation. This information is essential with regards to the heterogeneity of the neutron transfer to the organism. This is a very important parameter for individual dose reconstruction from the surface to the deeper tissues. This evaluation is essential to the adapted management of irradiated victims by specialized medical staff. Comparison exercises between clinical biochemistry laboratories from French sites (the CEA and COGEMA) and from the IRSN were carried out to validate the measurement of (32)P activity in hair and to improve the techniques used to perform this examination. Hair was placed on a phantom and was irradiated at different doses in the SILENE reactor (Valduc, France). Different parameters were tested: variation of hair type, minimum weight of hair sample, hair wash before measurement, delivery period of results, and different irradiation configurations. The results obtained in these comparison exercises by the different laboratories showed an excellent correlation. This allowed the assessment of a dose-activity relationship and confirmed the feasibility and the interest of (32)P measurement in hair following fast neutron irradiation.

  13. Crystalline hydroxyapatite coatings synthesized under hydrothermal conditions on modified titanium substrates.

    Science.gov (United States)

    Suchanek, Katarzyna; Bartkowiak, Amanda; Gdowik, Agnieszka; Perzanowski, Marcin; Kąc, Sławomir; Szaraniec, Barbara; Suchanek, Mateusz; Marszałek, Marta

    2015-06-01

    Hydroxyapatite coatings were successfully produced on modified titanium substrates via hydrothermal synthesis in a Ca(EDTA)(2-) and (NH4)2HPO4 solution. The morphology of modified titanium substrates as well as hydroxyapatite coatings was studied using scanning electron microcopy and phase identification by X-ray diffraction, and Raman and FTIR spectroscopy. The results show that the nucleation and growth of hydroxyapatite needle-like crystals with hexagonal symmetry occurred only on titanium substrates both chemically and thermally treated. No hydroxyapatite phase was detected on only acid etched Ti metal. This finding demonstrates that only a particular titanium surface treatment can effectively induce the apatite nucleation under hydrothermal conditions. Copyright © 2015 Elsevier B.V. All rights reserved.

  14. The intrinsic gettering in neutron irradiation Czochralski-silicon

    CERN Document Server

    Li Yang Xian; Niu Ping Juan; Liu Cai Chi; Xu Yue Sheng; Yang Deren; Que Duan Lin

    2002-01-01

    The intrinsic gettering in neutron irradiated Czochralski-silicon is studied. The result shows that a denuded zone at the surface of the neutron irradiated Czochralski-silicon wafer may be formed through one-step short-time annealing. The width of the denuded zone is dependent on the annealing temperature and the dose of neutron irradiation, while it is irrelated to the annealing time in case the denuded zone is formed. The authors conclude that the interaction between the defects induced by neutron irradiation and the oxygen in the silicon accelerates the oxygen precipitation in the bulk, and becomes the dominating factor of the quick formation of intrinsic gettering. It makes the effect of thermal history as the secondary factor

  15. A neutron irradiator applied to cancer treatment

    International Nuclear Information System (INIS)

    Campos, Tarcisio P.R.; Andrade, Ana P. de

    2000-01-01

    Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

  16. Detection and measurement of neutron-irradiated gemstones

    International Nuclear Information System (INIS)

    Bunnak, S.; Jerachanchai, S.; Chinudomsub, K.; Saiyut, K.

    1990-01-01

    Color enhance gemstone, neutron-irradiated topaz, was analyzed by gamma spectrometry for examining characteristic and activity. Topaz was irradiated in the wet-tube facility of the Research Reactor TRR/1 which neutron fluence is 2.52x10 17 neutron per square centimeter. After 100 days of decay, topaz was sampling to the qualitative and quantitative analysis using multichannel analyzer of Nuclear Data Model ND65 and hyper pure germanium detector. Calculation and evaluation were done by microcomputer IBM/PC 640 KB RAM. The qualitative analysis showed that the neutron-irradiated topaz has 2 major isotopes, i.e., Ta-182 and Sc-46. Quantitative activity was compared with reference standard source Eu-152 (NBS) and the results were shown in the table 1. The Health Physics Division, OAEP, inspected on 6240.9 gm of the neutron-irradiated topaz using standard release limit 2 nCi/gm (74 Bq/gm). It was found that only 423.9 gm out of the total amount were over the standard release limit

  17. Polarizing neutron by light-irradiated graphene

    International Nuclear Information System (INIS)

    Peng, Feng

    2015-01-01

    We study the spin orientation of the neutron scattered by light-irradiated graphene and calculate the average value of spin z-component of the neutron in terms of a generating functional technique. Our calculation results indicate that there is a remarkable neutron polarization effect when a neutron penetrates graphene irradiated by a circularly polarized light. We analyse the dynamical source of generating this effect from the aspect of photon-mediated interaction between the neutron spin and valley pseudospin. By comparing with the polarization induced by a magnetic field, we find that this polarization may be equivalent to the one led by a magnetic field of several hundred Teslas if the photon frequency is in the X-ray frequency range. This provides an approach of polarizing neutrons. (copyright 2015 by WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  18. Attenuation capability of low activation-modified high manganese austenitic stainless steel for fusion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, M.M. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El-kameesy, S.U.; El-Fiki, S.A. [Physics Department, Faculty of Science, Ain Shams University, Cairo (Egypt); Ghali, S.N. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El Shazly, R.M. [Physics Department, Faculty of Science, Al-Azhar University, Cairo (Egypt); Saeed, Aly, E-mail: aly_8h@yahoo.com [Nuclear Power station Department, Faculty of Engineering, Egyptian-Russian University, Cairo (Egypt)

    2016-11-15

    Highlights: • Improvement stainless steel alloys to be used in fusion reactors. • Structural, mechanical, attenuation properties of investigated alloys were studied. • Good agreement between experimental and calculated results has been achieved. • The developed alloys could be considered as candidate materials for fusion reactors. - Abstract: Low nickel-high manganese austenitic stainless steel alloys, SSMn9Ni and SSMn10Ni, were developed to use as a shielding material in fusion reactor system. A standard austenitic stainless steel SS316L was prepared and studied as a reference sample. The microstructure properties of the present stainless steel alloys were investigated using Schaeffler diagram, optical microscopy, and X-ray diffraction pattern. Mainly, an austenite phase was observed for the prepared stainless steel alloys. Additionally, a small ferrite phase was observed in SS316L and SSMn10Ni samples. The mechanical properties of the prepared alloys were studied using Vickers hardness and tensile tests at room temperature. The studied manganese stainless steel alloys showed higher hardness, yield strength, and ultimate tensile strength than SS316L. On the other hand, the manganese stainless steel elongation had relatively lower values than the standard SS316L. The removal cross section for both slow and total slow (primary and those slowed down in sample) neutrons were carried out using {sup 241}Am-Be neutron source. Gamma ray attenuation parameters were carried out for different gamma ray energy lines which emitted from {sup 60}Co and {sup 232}Th radioactive sources. The developed manganese stainless steel alloys had a higher total slow removal cross section than SS316L. While the slow neutron and gamma rays were nearly the same for all studied stainless steel alloys. From the obtained results, the developed manganese stainless steel alloys could be considered as candidate materials for fusion reactor system with low activation based on the short life

  19. A histologic analysis of the effects of stainless steel and titanium implants adjacent to tendons: an experimental rabbit study.

    Science.gov (United States)

    Nazzal, Adam; Lozano-Calderón, Santiago; Jupiter, Jesse B; Rosenzweig, Jaime S; Randolph, Mark A; Lee, Sang Gil P

    2006-09-01

    The current trend is to treat distal radius fractures with open reduction and internal fixation with either titanium or stainless steel plates. Both provide stable fixation; however, there is minimal evidence concerning the soft-tissue response to these materials. Our objective was to evaluate the response of adjacent extensor tendons to titanium and stainless steel in a rabbit in vivo model and to evaluate the influence of time. Forty rabbits were divided into 5 groups of 8 rabbits each. Groups I and II had unilateral osteotomy of the distal radius followed by dorsal fixation with titanium and stainless steel plates, respectively. Groups III and IV had fixation with titanium and stainless steel, respectively, but without osteotomy. Group V had surgical dissection without osteotomy or plates. Two animals per group were killed at 1, 4, 12, and 24 weeks. The specimens (distal radius, plate, overlying soft tissue, and extensor tendon) were harvested en bloc for histologic analysis. For interface preservation between implant and tissues the specimens were embedded in methylmethacrylate, sectioned, and stained with hematoxylin-eosin. Histologic analysis showed a fibrous tissue layer formed over both implants between the plate and the overlying extensor tendons in the groups treated with plating independently of the material and the presence or absence of osteotomy. This fibrous layer contained the majority of debris. Metallic particles were not observed in the tendon or muscle substance of any animals; however, they were visualized in the tenosynovium. Hematoxylin-eosin-stained sections of groups I through IV showed proliferative fibroblasts and metallic particles; however, this layer was not observed in group V. Statistical analysis did not show differences between the groups regarding the number of cells or metallic particles. Our results indicate that both implants generated adjacent reactive inflammatory tissue and particulate debris. There was no difference in cell

  20. Needs of in-situ materials testing under neutron irradiation

    International Nuclear Information System (INIS)

    Noda, K.; Hishinuma, A.; Kiuchi, K.

    1989-01-01

    Under neutron irradiation, the component atoms of materials are displaced as primary knock-on atoms, and the energy of the primary knock-on atoms is consumed by electron excitation and nuclear collision. Elementary irradiation defects accumulate to form damage structure including voids and bubbles. In situ test under neutron irradiation is necessary for investigating into the effect of irradiation on creep behavior, the electric properties of ceramics, transport phenomena and so on. The in situ test is also important to investigate into the phenomena related to the chemical reaction with environment during irradiation. Accelerator type high energy neutron sources are preferable to fission reactors. In this paper, the needs and the research items of in situ test under neutron irradiation using a D-Li stripping type high energy neutron source on metallic and ceramic materials are described. Creep behavior is one of the most important mechanical properties, and depends strongly on irradiation environment, also it is closely related to microstructure. Irradiation affects the electric conductibity of ceramics and also their creep behavior. In this way, in situ test is necessary. (K.I.)

  1. Structure and adhesive ability of (TiAl)N deposit on stainless steel

    International Nuclear Information System (INIS)

    Shiryaev, S.A.; Mitin, A.V.; Atamanov, M.V.; Moskovkin, P.G.; Guseva, M.I.; Mitin, V.S.

    2002-01-01

    The (TiAl)N coating on the stainless steel is obtained through the method of atomization in a magnetron with a mosaic cathode. The synthesized coating consists of the mixture of amorphous and nanocrystalline units of the titanium, aluminium nitrides and the titanium-aluminium alloy. The preliminary implantation of the nitrogen ions or irradiation, leading to the martensitic transformations, create the compressive strain in the near-the-surface layers of the sublayer and improve the adhesion. The model, describing the film adhesion and accounting for the strains therein and in the sublayer, is proposed [ru

  2. Observation of neutron bursts in saturation of titanium with deuterium by means of D2O electrolysis

    International Nuclear Information System (INIS)

    Artyukhov, V.I.; Bystritskij, V.M.; Gilev, A.I.

    1991-01-01

    The paper describes a correlation experiment on investigation of low-temperature nuclear dd-fusion during saturation of titanium with deuterium through electrolysis of heavy water D 2 O. The experiments with cathodes of chemically pure titanium and of titanium coated with a 0.4μm nickel layer (mass of titanium 26 g) were carried out. Emission of neutrons in the form of separate bursts was observed in the experiments with the nickel-coated cathode. The neutron emission density in the burst was found to be I n =(3.6±0.9)x10 4 s -1 . 17 refs.; 6 figs

  3. Endodontic complications of root canal therapy performed by dental students with stainless-steel K-files and nickel-titanium hand files.

    Science.gov (United States)

    Pettiette, M T; Metzger, Z; Phillips, C; Trope, M

    1999-04-01

    Straightening of curved canals is one of the most common procedural errors in endodontic instrumentation. This problem is commonly encountered when dental students perform molar endodontics. The purpose of this study was to compare the effect of the type of instrument used by these students on the extent of straightening and on the incidence of other endodontic procedural errors. Nickel-titanium 0.02 taper hand files were compared with traditional stainless-steel 0.02 taper K-files. Sixty molar teeth comprised of maxillary and mandibular first and second molars were treated by senior dental students. Instrumentation was with either nickel-titanium hand files or stainless-steel K-files. Preoperative and postoperative radiographs of each tooth were taken using an XCP precision instrument with a customized bite block to ensure accurate reproduction of radiographic angulation. The radiographs were scanned and the images stored as TIFF files. By superimposing tracings from the preoperative over the postoperative radiographs, the degree of deviation of the apical third of the root canal filling from the original canal was measured. The presence of other errors, such as strip perforation and instrument breakage, was established by examining the radiographs. In curved canals instrumented by stainless-steel K-files, the average deviation of the apical third of the canals was 14.44 degrees (+/- 10.33 degrees). The deviation was significantly reduced when nickel-titanium hand files were used to an average of 4.39 degrees (+/- 4.53 degrees). The incidence of other procedural errors was also significantly reduced by the use of nickel-titanium hand files.

  4. Surface characterisation and electrochemical behaviour of porous titanium dioxide coated 316L stainless steel for orthopaedic applications

    International Nuclear Information System (INIS)

    Nagarajan, S.; Rajendran, N.

    2009-01-01

    Porous titanium dioxide was coated on surgical grade 316L stainless steel (SS) and its role on the corrosion protection and enhanced biocompatibility of the materials was studied. X-ray diffraction analysis (XRD), atomic force microscopy (AFM), Fourier transform infrared (FTIR) spectroscopy, scanning electron microscopy (SEM) and energy dispersive X-ray analysis (EDAX) were carried out to characterise the surface morphology and also to understand the structure of the as synthesised coating on the substrates. The corrosion behaviour of titanium dioxide coated samples in simulated body fluid was evaluated using polarisation and impedance spectroscopy studies. The results reveal that the titanium dioxide coated 316L SS exhibit a higher corrosion resistance than the uncoated 316L SS. The titanium dioxide coated surface is porous, uniform and also it acts as a barrier layer to metallic substrate and the porous titanium dioxide coating induces the formation of hydroxyapatite layer on the metal surface.

  5. DNA-repair after irradiation of cells with gamma-rays and neutrons

    International Nuclear Information System (INIS)

    Altmann, H.

    1975-11-01

    The structural alterations of calf thymus DNA produced by neutron or gamma irradiation were observed by absorption spectra, sedimentation rate and viscosity measurements. Mixed neutron-gamma irradiation produced fewer single and double strand breaks compared with pure gamma irradiation. RBE-values for mixed neutron-gamma radiation were less than 1, and DNA damage decreased with increasing neutron dose rate. Repair processes of DNA occuring after irradiation were measured in mouse spleen suspensions and human lymphocytes using autoradiographic methods and gradient centrifugations. The number of labelled cells was smaller after mixed neutron-gamma irradiation than after gamma irradiation. The rejoining of strand breaks in alkaline and neutral sucrose was more efficient after gamma irradiation than after mixed neutron-gamma irradiation. Finally, the effect of detergents Tween 80 and Nonident P40 on unscheduled DNA synthesis was studied by autoradiography after mixed neutron-gamma irradiation (Dn=5 krad). The results showed that the DNA synthesis was inhibited by detergent solutions of 0.002%

  6. Superfluid He testing of titanium-stainless steel transitions fabricated by explosive welding

    International Nuclear Information System (INIS)

    Budagov, Yu.; Sabirov, B.; Shirkov, G.

    2009-01-01

    An experimental setup was constructed to test in liquid He bimetallic (titanium-stainless steel) tube joints which were manufactured by an explosive welding method. The leak levels of the samples tested at room temperature 7.5·10 -10 and 7.5·10 -9 Torr·1/s at 77 K, correspondingly, measured at FNAL (Batavia, USA) after the thermocycling have coincided with the earlier results obtained at JINR (Dubna, Russia) and INFN (Pisa, Italy) data for the same samples. For the liquid helium test the tubes were welded in pairs by their titanium ends. At the room temperature the leak level of the three tested samples was 4.9·10 -10 Torr·l/s. At the first cryogenic tests (4-6 K) one of the samples manifested a leak. The investigation will be continued since the explosive welding seems to be a very perspective new generation technology

  7. Study of boron carbide evolution under neutron irradiation

    International Nuclear Information System (INIS)

    Simeone, D.

    1999-01-01

    Owing to its high neutron efficiency, boron carbide (B 4 C) is used as a neutron absorber in control rods of nuclear plants. Its behaviour under irradiation has been extensively studied for many years. It now seems clear that brittleness of the material induced by the 10 B(n,α) 7 Li capture reaction is due to penny shaped helium bubbles associated to a high strain field around them. However, no model explains the behaviour of the material under neutron irradiation. In order to build such a model, this work uses different techniques: nuclear microprobe X-ray diffraction profile analysis and Raman and Nuclear Magnetic Resonance Spectroscopy to present an evolution model of B 4 C under neutron irradiation. The use of nuclear reactions produced by a nuclear microprobe such as the 7 Li(p,p'γ) 7 Li reaction, allows to measure lithium profile in B 4 C pellets irradiated either in Pressurised Water Reactors or in Fast Breeder Reactors. Examining such profiles enables us to describe the migration of lithium atoms out of B 4 C materials under neutron irradiation. The analysis of X-ray diffraction profiles of irradiated B 4 C samples allows us to quantify the concentrations of helium bubbles as well as the strain fields around such bubbles.Furthermore Raman spectroscopy studies of different B 4 C samples lead us to propose that under neutron irradiation. the CBC linear chain disappears. Such a vanishing of this CBC chain. validated by NMR analysis, may explain the penny shaped of helium bubbles inside irradiated B 4 C. (author)

  8. Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

    International Nuclear Information System (INIS)

    Brown, B.S.; Blewitt, T.H.

    1982-01-01

    The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

  9. Assessment of antibacterial and cytotoxic effects of orthodontic stainless steel brackets coated with different phases of titanium oxide: An in-vitro study.

    Science.gov (United States)

    Baby, Roshen Daniel; Subramaniam, Siva; Arumugam, Ilakkiya; Padmanabhan, Sridevi

    2017-04-01

    Our objective was to assess the antibacterial and cytotoxic effects of orthodontic stainless steel brackets coated with different phases of photocatalytic titanium oxide. From a total sample of 115 brackets, 68 orthodontic stainless steel brackets were coated with titanium oxide using a radiofrequency magnetron sputtering machine. The coated brackets were then converted into 34 each of the anatase and rutile phases of titanium oxide. These brackets were subdivided into 4 groups for antibacterial study and 3 groups for cytotoxicity study. Brackets for the antibacterial study were assessed against the Streptococcus mutans species using microbiologic tests. Three groups for the cytotoxicity study were assessed using the thiazolyl tetrazolium bromide assay. The antibacterial study showed that both phases were effective, but the rutile phase of photocatalytic titanium oxide had a greater bactericidal effect than did the anatase phase. The cytotoxicity study showed that the rutile phase had a greater decrease in viability of cells compared with the anatase phase. It is recommended that orthodontic brackets be coated with the anatase phase of titanium oxide since they exhibited a significant antibacterial property and were only slightly cytotoxic. Copyright © 2016 American Association of Orthodontists. Published by Elsevier Inc. All rights reserved.

  10. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO{sub 3} and H{sub 2}O{sub 2} solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area).

  11. Reactor Materials Program electrochemical potential measurements by ORNL with unirradiated and irradiated stainless steel specimens

    International Nuclear Information System (INIS)

    Baumann, E.W.; Caskey, G.R. Jr.

    1993-07-01

    Effect of irradiation of stainless steel on electrochemical potential (ECP) was investigated by measurements in dilute HNO 3 and H 2 O 2 solutions, conditions simulating reactor moderator. The electrodes were made from unirradiated/irradiated, unsensitized/sensitized specimens from R-reactor piping. Results were inconclusive because of budgetary restrictions. The dose rate may have been too small to produce a significant radiolytic effect. Neither the earlier CERT corrosion susceptibility tests nor the present ECP measurements showed a pronounced effect of irradiation on susceptibility of the stainless steel to IGSCC; this is confirmed by the absence in the stainless steel of the SRS reactor tanks (except for the C Reactor tank knuckle area)

  12. Impact of neutron irradiation on thermal helium desorption from iron

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Xunxiang, E-mail: hux1@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Field, Kevin G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Taller, Stephen [University of Michigan, Ann Arbor, MI 48109 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Wirth, Brian D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States)

    2017-06-15

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 10{sup 18} He/m{sup 2}. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  13. Radiation damage of pixelated photon detector by neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Isamu [KEK, 1-1 Oho Tsukuba 305-0801 (Japan)], E-mail: isamu.nakamura@kek.jp

    2009-10-21

    Radiation Damage of Pixelated Photon Detector by neutron irradiation is reported. MPPC, one of PPD or Geiger-mode APD, developed by Hamamatsu Photonics, is planned to be used in many high energy physics experiments. In such experiments radiation damage is a serious issue. A series of neutron irradiation tests is performed at the Reactor YAYOI of the University of Tokyo. MPPCs were irradiated at the reactor up to 10{sup 12}neutron/cm{sup 2}. In this paper, the effect of neutron irradiation on the basic characteristics of PPD including gain, noise rate, photon detection efficiency is presented.

  14. Induced defects in neutron irradiated GaN single crystals

    International Nuclear Information System (INIS)

    Park, I. W.; Koh, E. K.; Kim, Y. M.; Choh, S. H.; Park, S. S.; Kim, B. G.; Sohn, J. M.

    2005-01-01

    The local structure of defects in undoped, Si-doped, and neutron irradiated free standing GaN bulk crystals, grown by hydride vapor phase epitaxy, has been investigated by employing Raman scattering and cathodoluminescence. The GaN samples were irradiated to a dose of 2 x 10 17 neutrons in an atomic reactor at Korea Atomic Energy Research Institute. There was no appreciable change in the Raman spectra for undoped GaN samples before and after neutron irradiation. However, a forbidden transition, A 1 (TO) mode, appeared for a neutron irradiated Si-doped GaN crystal. Cathodoluminescence spectrum for the neutron irradiated Si-doped GaN crystal became much more broadened than that for the unirradiated one. The experimental results reveal the generation of defects with locally deformed structure in the wurtzite Si-doped GaN single crystal

  15. Design, Construction, and Modeling of a 252Cf Neutron Irradiator

    Directory of Open Access Journals (Sweden)

    Blake C. Anderson

    2016-01-01

    Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

  16. Irradiation hardening and localized deformation of neutron-irradiated α-iron single crystals

    International Nuclear Information System (INIS)

    Mughrabi, H.; Stroehle, D.; Wilkens, M.

    1981-01-01

    The early yielding behaviour of neutron-irradiated α iron single crystals orientated for single slip was investigated as a function of neutron dose. In the range of neutron doses between approx. equal to 10 18 and approx. equal to 10 19 n/cm 2 , the irradiation hardening increment was found to be almost constant. Qualitative modifications of this behaviour were observed in the case of predeformed specimens. The localized deformation of the neutron-irradiated specimens by dislocation channelling was investigated by slip-line observations, transmission electron microscopy and X-ray topography. A model of localized deformation is proposed in order to explain the development of the observed asymmetric dislocation double layers which bound the channels and transmit characteristic misorientations. (orig.)

  17. Technology development on production of test specimens from irradiated capsule outer-tube and mechanical evaluation test of stainless steel with high dose carried out by the technology

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya

    2008-03-01

    The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.9x10 26 n/m 2 at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer-tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed with in-water high temperature condition, less than 288degC. The result shows that the residual elongation is 18% at 150degC and 13% at 288degC. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility. (author)

  18. Neutron irradiation therapy machine

    International Nuclear Information System (INIS)

    1980-01-01

    Conventional neutron irradiation therapy machines, based on the use of cyclotrons for producing neutron beams, use a superconducting magnet for the cyclotron's magnetic field. This necessitates complex liquid He equipment and presents problems in general hospital use. If conventional magnets are used, the weight of the magnet poles considerably complicates the design of the rotating gantry. Such a therapy machine, gantry and target facilities are described in detail. The use of protons and deuterons to produce the neutron beams is compared and contrasted. (U.K.)

  19. Compared behaviour of stainless steel and titanium tests lines for a L.H.H. grill launcher

    International Nuclear Information System (INIS)

    Rey, G.; Briand, P.; David, C.; Lamboley, G.; Moulin, D.; Tonon, G.

    1982-09-01

    The choice of the material as first wall of a grill is of major importance. It must satisfy eventually conficting demands: (i) mean secondary emission coefficient low enough to avoid multipactor effect, (ii) low gas absorption and desorption and (iii) easy manufacturing. Then two test lines, one made of pure titanium, the other made of 304 L stainless steel have been tested at RF power greater than 300 kW during 150 ms at 1,25 GHz. As reported in a previous paper, 360 kW have been transmitted in the titanium test line without or with a very slight change in phase between incident an transmitted wave ( 0 ) after application of low pressure argon glow discharges. We now report the first results obtained on a stainless steel test line after the use of the same conditioning procedure (baking and argon glow discharges) and the effects of a static magnetic field superimposed in the reduced size sections of the line (B parallel to the large dimension of the waveguide)

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

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu; Kimura, Hideo

    1985-01-01

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

  1. Effect of neutron irradiation on vitreous carbon

    International Nuclear Information System (INIS)

    Kurolenkin, E.I.; Virgil'ev, Yu.S.; Chugunova, T.K.

    1989-01-01

    The change in mass (m), volume (V), specific electric resistance (ρ), coefficient of linear thermal expansion (α), dynamic elasticity modulus (E), and limit of bending strength (σ) of vitreous carbon are studied upon neutron irradiation. Samples for study were two forms of vitreous carbon obtained by hardening thermally reactive polymers at 900-1,000 degree K. Phenol-formaldehyde (bakelite lacquer A, Bakelite A) and furfural-phenol-formaldehyde (FM-2) resin were used. They were irradiated in the experimental water - water VVR-M reactor between 360-1,030 degree K. The maximal neutron flux was 1.65·10 21 neut/cm 2 . Neutron irradiation of vitreous carbon led to its shrinkage and accompanied weakening. Shrinkage and weakening of vitreous carbon was decreased with an increase of treatment and irradiation temperatures

  2. Thermal creep properties of alloy D9 stainless steel and 316 stainless steel fuel clad tubes

    International Nuclear Information System (INIS)

    Latha, S.; Mathew, M.D.; Parameswaran, P.; Bhanu Sankara Rao, K.; Mannan, S.L.

    2008-01-01

    Uniaxial thermal creep rupture properties of 20% cold worked alloy D9 stainless steel (alloy D9 SS) fuel clad tubes for fast breeder reactors have been evaluated at 973 K in the stress range 125-250 MPa. The rupture lives were in the range 90-8100 h. The results are compared with the properties of 20% cold worked type 316 stainless steel (316 SS) clad tubes. Alloy D9 SS were found to have higher creep rupture strengths, lower creep rates and lower rupture ductility than 316 SS. The deformation and damage processes were related through Monkman Grant relationship and modified Monkman Grant relationship. The creep damage tolerance parameter indicates that creep fracture takes place by intergranular cavitation. Precipitation of titanium carbides in the matrix and chromium carbides on the grain boundaries, dislocation substructure and twins were observed in transmission electron microscopic investigations of alloy D9 SS. The improvement in strength is attributed to the precipitation of fine titanium carbides in the matrix which prevents the recovery and recrystallisation of the cold worked microstructure

  3. Stress corrosion cracking of highly irradiated 316 stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, Morihito; Fukuya, Koji; Fujii, Katsuhiko; Nakajima, Nobuo; Furutani, Gen [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    Mechanical property tests, grain boundary (GB) composition analysis and slow strain rate test (SSRT) in simulated PWR primary water changing dissolved hydrogen (DH) and dissolved oxygen (DO) content were carried out on cold-worked (CW) 316 stainless steels which were irradiated to 1-8x10{sup 26} n/m{sup 2} (E>0.1 MeV) in a Japanese PWR in order to evaluate irradiation-assisted stress corrosion cracking (IASCC) susceptibility. Highly irradiated stainless steels were susceptible to intergranular stress corrosion cracking (IGSCC) in both hydrogenated water and oxygenated water and to intergranular cracking in inert gas atmosphere. IASCC susceptibility increased with increasing DH content (0-45 ccH{sub 2}/kgH{sub 2}O). Hydrogen content of the section containing fracture surface was higher than that of the section far from fracture surface. These results suggest that hydrogen would have an important role for IASCC. While mechanical property was saturated, GB segregation and IASCC susceptibility increased with an increase in fluence, suggesting that GB segregation would have a dominant role for an increase in IASCC susceptibility at this high fluence region. (author)

  4. Biological Effects of Neutron and Proton Irradiations. Vol. II. Proceedings of the Symposium on Biological Effects of Neutron Irradiations

    International Nuclear Information System (INIS)

    1964-01-01

    During recent years the interest in biological effects caused by neutrons has been increasing steadily as a result of the rapid development of neutron technology and the great number of neutron sources being used. Neutrons, because of their specific physical characteristics and biological effects, form a special type of radiation hazard but, at the same time, are a prospective tool for applied radiobiology. This Symposium, held in Brookhaven at the invitation of the United States Government from 7-11 October 1963, provided an opportunity for scientists to discuss the experimental information at present available on the biological action of neutrons and to evaluate future possibilities. It was a sequel to the Symposium on Neutron Detection, Dosimetry and Standardization, which was organized by the International Atomic Energy Agency in December 1962 at Harwell. The Symposium was attended by 128 participants from 17 countries and 6 international organizations. Fifty-four papers were presented. The following subjects were discussed in various sessions: (1) Dosimetry. Estimation of absorbed dose of neutrons in biological material. (2) Biological effects of high-energy protons. (3) Cellular and genetic effects. (4) Pathology of neutron irradiation, including acute and chronic radiation syndromes (mortality, anatomical and histological changes, biochemical and metabolic disturbances) and delayed consequences. (5) Relative biological effectiveness of neutrons evaluated by different biological tests. A Panel on Biophysical Considerations in Neutron Experimentation, with special emphasis on informal discussions, was organized during the Symposium. The views of the Panel are recorded in Volume II of the Proceedings. Many reports were presented on the important subject of the relative effectiveness of the biological action of neutrons, as well as on the general pathology of neutron irradiation and the cellular and genetic effects related to it. Three survey papers considered

  5. Visible Light Photocatalytic Properties of Modified Titanium Dioxide Nanoparticles via Aluminium Treatment

    Directory of Open Access Journals (Sweden)

    Dessy Ariyanti

    2016-03-01

    Full Text Available Titanium dioxide (TiO2 has gained much attentions for the last few decades due to its remarkable performance in photocatalysis and some other related properties. However, its wide bandgap (~3.2 eV can only absorb UV energy which is only ~5% of solar light spectrum. The objective of this research was to improve the photocatalytic activity of TiO2 by improving the optical absorption to the visible light range. Here, colored TiO2 nanoparticles range from light to dark grey were prepared via aluminium treatment at the temperatures ranging from 400 to 600 oC. The modified TiO2 is able to absorb up to 50% of visible light (400-700 nm and shows a relatively good photocatalytic activity in organic dye (Rhodamine B degradation under visible light irradiation compared with the commercial TiO2. Copyright © 2016 BCREC GROUP. All rights reserved Received: 10th November 2015; Revised: 7th January 2016; Accepted: 7th January 20 How to Cite: Ariyanti, D., Dong, J.Z., Dong, J.Y., Gao, W. (2016. Visible Light Photocatalytic Properties of Modified Titanium Dioxide Nanoparticles via Aluminium Treatment. Bulletin of Chemical Reaction Engineering & Catalysis, 11 (1: 40-47. (doi:10.9767/bcrec.11.1.414.40-47 Permalink/DOI: http://dx.doi.org/10.9767/bcrec.11.1.414.40-47

  6. Comparative study of two materials for dynamic hip screw during fall and gait loading: titanium alloy and stainless steel.

    Science.gov (United States)

    Taheri, Nooshin S; Blicblau, Aaron S; Singh, Manmohan

    2011-11-01

    Internal fixation with dynamic hip screw is a choice of treatment for hip fractures to stabilize a femoral fracture. Choosing the proper implant and its material has a great effect on the healing process and failure prevention. The purpose of this analysis was to assess biomechanical behavior of dynamic hip screw with two different materials implanted in the femur during fall and gait. A 3D finite element model of an intact femur and a 3D implant within the same femur were developed. A finite element analysis was carried out to establish the effect of load conditions and implant material properties on biomechanical behavior of the dynamic hip screw after internal fixation. Two load configurations are chosen: one simulating the stance phase of the normal gait cycle, and the other replicating a low-energy fall. The implanted femur was investigated with two different materials for the dynamic hip screw: stainless steel and titanium alloy. During stance, more stress is placed on the implanted femur compared with the intact femur. During a fall, the implanted femur is in a greater state of stress, which mostly occurs inside the dynamic hip screw. Titanium alloy decreases stress levels by an average of 40% compared with stainless steel. However, deformation is slightly reduced with a stainless steel dynamic hip screw during both load cases. After internal fixation, dynamic hip screw generates greater stresses within the implanted femur compared with the intact femur under the same loading conditions. A titanium alloy implant appears to undergo less stress from a low-energy fall compared with stainless steel and can be considered the preferred implant material. The critical parts of the dynamic hip screw are the forth distal screw and the plate.

  7. Radiation-induced instability of MnS precipitates and its possible consequences on irradiation-induced stress corrosion cracking of austenitic stainless steels

    International Nuclear Information System (INIS)

    Chung, H.M.; Sanecki, J.E.; Garner, F.A.

    1996-12-01

    Irradiation-assisted stress corrosion cracking (IASCC) is a significant materials issue for the light water reactor (LWR) industry and may also pose a problem for fusion power reactors that will use water as coolant. A new metallurgical process is proposed that involves the radiation-induced release into solution of minor impurity elements not usually thought to participate in IASCC. MnS-type precipitates, which contain most of the sulfur in stainless steels, are thought to be unstable under irradiation. First, Mn transmutes strongly to Fe in thermalized neutron spectra. Second, cascade-induced disordering and the inverse Kirkendall effect operating at the incoherent interfaces of MnS precipitates are thought to act as a pump to export Mn from the precipitate into the alloy matrix. Both of these processes will most likely allow sulfur, which is known to exert a deleterious influence on intergranular cracking, to re-enter the matrix. To test this hypothesis, compositions of MnS-type precipitates contained in several unirradiated and irradiated heats of Type 304, 316, and 348 stainless steels (SSs) were analyzed by Auger electron spectroscopy. Evidence is presented that shows a progressive compositional modification of MnS precipitates as exposure to neutrons increases in boiling water reactors. As the fluence increases, the Mn level in MnS decreases, whereas the Fe level increases. The S level also decreases relative to the combined level of Mn and Fe. MnS precipitates were also found to be a reservoir of other deleterious impurities such as F and O which could be also released due to radiation-induced instability of the precipitates

  8. Study of damages by neutron irradiation in lithium aluminates

    International Nuclear Information System (INIS)

    Palacios G, O.

    1999-01-01

    Lithium aluminates proposed to the production of tritium in fusion nuclear reactors, due to the thermal stability that they present as well as the behavior of the aluminium to the irradiation. As a neutron flux with profile (≅ 14 Mev) of a fusion reactor is not available. A irradiation experiment was designed in order to know the micro and nano structure damages produced by fast and thermal neutrons in two irradiation positions of the fusion nuclear reactor Triga Mark III: CT (Thermal Column) and SIFCA (System of Irradiation Fixed of Capsules). In this work samples of lithium aluminate were characterized by XRD (X-Ray Diffraction), TEM (Transmission Electron Microscopy) and SEM (Scanning Electron Microscopy). Two samples were prepared by two methods: a) coalition method and b) peroxide method. This characterization comprised original and irradiated samples. The irradiated sample amounted to 4 in total: one for each preparation method and one for each irradiation position. The object of this analysis was to correlate with the received neutron dose the damages suffered by the samples with the neutron irradiation during long periods (440 H), in their micro and nano structure aspects; in order to understand the changes as a function of the irradiation zone (with thermal and fast neutron flux) and the preparation methods of the samples and having as an antecedent the irradiation in SIFCA position by short times (2h). The obtained results are referred to the stability of γ -aluminate phase, under given conditions of irradiation and defined nano structure arrangement. They also refer to the proposals of growth mechanism and nucleation of new phases. The error associated with the measurement of neutron dose is also discussed. (Author)

  9. Neutron resistant irradiation alloy and usage thereof

    International Nuclear Information System (INIS)

    Okada, Osamu; Nakata, Kiyotomo; Kato, Takahiko.

    1997-01-01

    A neutron irradiation embrittlement-resistant alloy comprising a Ti alloy having an average grain size of 2μm or smaller and containing from 30 to 40wt% of Al is subjected to powder solidification and then to isothermal forging at a forging rate of from 50 to 80% at a temperature range of from 1150 to 1500K. Namely, since the Ti-Al type alloy comprises from 30 to 30wt% of Al, optionally, from 1 to 6% of Mn, from 0.1 to 0.5% of Si, from 4 to 16% of V and the balance of Ti, it has excellent specific strength, high durable temperature and excellent neutron irradiation resistance, and has ductility required as structural materials. Accordingly, if the Ti-Al type alloy excellent in embrittlement resistance to neutron irradiation dimensional stability of materials is applied to constitutional parts of a reactor core of a nuclear reactor and a thermonuclear reactor to be exposed under neutron irradiation, high reliability is provided and the amount of activated materials is reduced by improving the working life of the materials. (N.H.)

  10. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2008-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  11. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2009-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  12. Investigation of the antibiofilm capacity of peptide-modified stainless steel.

    Science.gov (United States)

    Cao, Pan; Li, Wen-Wu; Morris, Andrew R; Horrocks, Paul D; Yuan, Cheng-Qing; Yang, Ying

    2018-03-01

    Biofilm formation on surfaces is an important research topic in ship tribology and medical implants. In this study, dopamine and two types of synthetic peptides were designed and attached to 304 stainless steel surfaces, aiming to inhibit the formation of biofilms. A combinatory surface modification procedure was applied in which dopamine was used as a coupling agent, allowing a strong binding ability with the two peptides. X-ray photoelectron spectroscopy (XPS), elemental analysis, contact angle measurement and surface roughness test were used to evaluate the efficiency of the peptide modification. An antibiofilm assay against Staphylococcus aureus was conducted to validate the antibiofilm capacity of the peptide-modified stainless steel samples. XPS analysis confirmed that the optimal dopamine concentration was 40 µg ml -1 in the coupling reaction. Element analysis showed that dopamine and the peptides had bound to the steel surfaces. The robustness assay of the modified surface demonstrated that most peptide molecules had bound on the surface of the stainless steel firmly. The contact angle of the modified surfaces was significantly changed. Modified steel samples exhibited improved antibiofilm properties in comparison to untreated and dopamine-only counterpart, with the peptide 1 modification displaying the best antibiofilm effect. The modified surfaces showed antibacterial capacity. The antibiofilm capacity of the modified surfaces was also surface topography sensitive. The steel sample surfaces polished with 600# sandpaper exhibited stronger antibiofilm capacity than those polished with other types of sandpapers after peptide modification. These findings present valuable information for future antifouling material research.

  13. Void shrinkage in stainless steel during high energy electron irradiation

    International Nuclear Information System (INIS)

    Singh, B.N.; Foreman, A.J.E.

    1976-03-01

    During irradiation of thin foils of an austenitic stainless steel in a high voltage electron microscope, steadily growing voids have been observed to suddenly shrink and disappear at the irradiation temperature of 650 0 Cthe phenomenon has been observed in specimens both with and withoutimplanted helium. Possible mechanisms for void shrinkage during irradiation are considered. It is suggested that the dislocation-pipe-diffusion of vacancies from or of self-interstitial atoms to the voids can explain the shrinkage behaviour of voids observed during our experiments. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Characteristics of modified martensitic stainless steel surfaces under tribocorrosion conditions

    International Nuclear Information System (INIS)

    Rozing, Goran; Marusic, Vlatko; Alar, Vesna

    2017-01-01

    Stainless steel samples were tested in the laboratory and under real conditions of tribocorrosion wear. Electrochemical tests were also carried out to verify the corrosion resistance of modified steel surfaces. Metallographic analysis and hardness testing were conducted on stainless steel samples X20Cr13 and X17CrNi16 2. The possibilities of applications of modified surfaces of the selected steels were investigated by testing the samples under real wear conditions. The results have shown that the induction hardened and subsequently nitrided martensitic steels achieved an average wear resistance of up to three orders of magnitude higher as compared to the delivered condition.

  16. Characteristics of modified martensitic stainless steel surfaces under tribocorrosion conditions

    Energy Technology Data Exchange (ETDEWEB)

    Rozing, Goran [Osijek Univ. (Croatia). Chair of Mechanical Engineering; Marusic, Vlatko [Osijek Univ. (Croatia). Dept. of Engineering Materials; Alar, Vesna [Zagreb Univ. (Croatia). Dept. Materials

    2017-04-01

    Stainless steel samples were tested in the laboratory and under real conditions of tribocorrosion wear. Electrochemical tests were also carried out to verify the corrosion resistance of modified steel surfaces. Metallographic analysis and hardness testing were conducted on stainless steel samples X20Cr13 and X17CrNi16 2. The possibilities of applications of modified surfaces of the selected steels were investigated by testing the samples under real wear conditions. The results have shown that the induction hardened and subsequently nitrided martensitic steels achieved an average wear resistance of up to three orders of magnitude higher as compared to the delivered condition.

  17. A neutron irradiator to perform nuclear activation

    International Nuclear Information System (INIS)

    Zamboni, C. B.; Zahn, G.S.; Figueredo, A. M. G.; Madi, T. F.; Yoriyaz, H.; Lima, R. B.; Shtejer, K.; Dalaqua Jr, L.

    2001-01-01

    The development of appropriate nuclear instrumentation to perform neutron activation analyze (NAA), using thermal and fast neutrons, can be useful to investigate materials outside the reactor premises. Considering this fact, a small size neutron irradiator prototype was developed at IPEN facilities (Instituto de Pesquisas Energeticas e Nucleares - Brazil). Basically, this prototype consists of a cylinder of 1200 mm long and 985 mm diameter (filled with paraffin) with two Am-Be sources (600GBq each) arranged in the longitudinal direction of its geometric center. The material to be irradiated is positioned at a radial direction of the cylinder between the two Am-Be sources. The main advantage of this irradiator is a very stable neutron flux eliminating the use of standard material (measure of the induced activity in the sample by comparative method). This way the process became agile, practical and economic, but quantities at mg levels of samples are necessary to achieve good sensitivity, when the material has a low microscopy neutron cross section. As fast and thermal neutron can be used, the flux distribution, for both, were calculated and the prototype performance is discussed

  18. Hydriding and neutron irradiation in zircaloy-4

    International Nuclear Information System (INIS)

    Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

    2003-01-01

    The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

  19. Neutron irradiation effects in pressure vessel steels and weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  20. A comparative evaluation of metallurgical properties of stainless steel and TMA archwires with timolium and titanium niobium archwires - An in vitro study

    Directory of Open Access Journals (Sweden)

    Vijayalakshmi R

    2009-01-01

    Full Text Available Objectives: This study aims to evaluate and compare the mechanical and metallurgical properties of stainless steel and titanium molybdenum alloy (TMA archwires, with recently introduced timolium and titanium niobium arch wires. Materials and Methods: Archwires were categorized into four groups (group I to IV with 10 samples in each group. They were evaluated for tensile strength, yield strength, modulus of elasticity, load deflection, frictional properties and weld characteristics. Results: The results were statistically analyzed using ANOVA test and it indicated that stainless steel has high strength, high stiffness and low friction compared to other arch wires, thereby proving that it is the best choice for both sliding as well as frictionless retraction mechanics. TMA with its high formability, low stiffness and low load deflection property is suited to apply consistent force in malaligned teeth but, high friction limits its use in retraction only with loop mechanics. Conclusion: Timolium possesses comparatively low stiffness, better strength and behaves as an intermediate between stainless steel and TMA and hence can be tried for almost all clinical situations. Low springback and high formability of titanium-niobium archwire allows creation of finishing bends and thus it can be used as finishing archwire.

  1. Minimization of the blank values in the neutron activation analysis of biological samples considering the whole procedure

    International Nuclear Information System (INIS)

    Lux, F.; Bereznai, T.; Haeberlin, T.H.

    1986-01-01

    In our determination of trace element contents of animal tissue by neutron activation analysis in the course of structure-activity relationship studies on platinum containing cancer drugs and wound healing we have tried to minimize the blank values that are caused by different sources of contamination during surgery, sampling and the activation analysis procedure. The following topics have been investigated: the abrasions from scalpels made of stainless steel, titanium or quartz; the type of surgery; the homogenisation of the samples before irradiation by use of a ball milll; the surface contaminations of the quartz ampoules that pass into the digestion solution of the irradiated samples. The appropriate measures taken in order to reduce the blank values are described. The results of analyses performed under these conditions indicate the effectiveness of the given measures, especially shown by the low values obtained for the chromium contents of the analysed muscle samples. (author)

  2. Austenitic stainless steel bulk property considerations for fusion reactors

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1979-04-01

    The bulk properties of annealed 304, 316, and 20% cold-worked 316 stainless steels are evaluated for the temperature and radiation conditions expected in a near-term fusion reactor. Of interest are the thermophysical properties, void swelling produced by neutron radiaion, and the tensile, creep, and fatigue properties before and after irradiation

  3. Damages to gladiolu corm caused by fast neutron irradiation

    International Nuclear Information System (INIS)

    Zhang Zhiwei; Wang Dan; Zhang Dongxue; Zheng Chun

    2007-01-01

    Gladiolus corms were irradiated to 100-500kGy by fast neutrons in the CFBR-II pulsed reactor, Scanning electron microscope images of the irradiated samples revealed significant radiation damages to the gladiolus corms, and the mutagenic effects were studied by SDS-polyacrylamide gel electrophoresis (SDS-PAGE). Within the dose range, radiation damage to the corm increased with the dose, with corm epidermis of the samples irradiated in vertical incidence being more serious than those irradiated in side-incidence to the same dose. Biological characters were investigated via field experiments, and the bands of protein subunit were analyzed by SDS-PAGE. The results showed that the fast neutrons irradiation inhibited growth of M1 generation seedling significantly. Protein expression was obviously inhibited by the irradiation. The study indicates that fast neutron induction is an effective way for gladiolus breeding. And the results may lay a foundation for studies on fast neutron mutation breeding. (authors)

  4. Fuel performance computer code simulation of steady-state and transient regimes of the stainless steel fuel rods

    International Nuclear Information System (INIS)

    Gomes, Daniel de Souza

    2014-01-01

    The immediate cause of the accident at the Fukushima Daiichi nuclear plant in March 2011 was the meltdown of the reactor core. During this process, the zirconium cladding of the fuel reacts with water, producing a large amount of hydrogen. This hydrogen, combined with volatile radioactive materials leaked from the containment vessel and entered the building of the reactor, resulting in explosions. In the past, stainless steel was used as the coating in many pressurized water reactors (PWR) under irradiation and their performance was excellent, however, the stainless steel was replaced by a zirconium-based alloy as a coating material mainly due to its lower section shock-absorbing neutrons. Today, the stainless steel finish appears again as a possible solution for security issues related to the explosion and hydrogen production. The objective of this thesis is to discuss the performance under irradiation of fuel rods using stainless steel as a coating material. The results showed that stainless steel rods exhibit lower temperatures and higher fuel pellet width of the gap - coating the coated rods Zircaloy and this gap does not close during the irradiation. The thermal performance of the two fuel rods is very similar, and the penalty of increased absorption of neutrons due to the use of stainless steel can be offset by the combination of a small increase in the enrichment of U- 235 and changes in the size of the spacing between the fuel rods. (author)

  5. Impurity effects in neutron-irradiated simple oxides: Implications for fusion devices

    International Nuclear Information System (INIS)

    Gonzalez, R.; Chen, Y.; Caceres, D.; Vergara, I.

    2006-01-01

    Radiation damage induced by neutron irradiation was studied in undoped MgO crystals and in MgO doped with either iron, hydrogen or lithium impurities. The oxygen-vacancy concentration produced by irradiation increases with neutron fluence. The net production rates resulting from irradiations with 14.8 MeV neutrons are about twice those produced by fission neutrons. In nominally pure crystals, the oxygen-vacancy concentration incurred by the fission-neutron irradiation is higher in crystals with a larger number of inherent impurities (such as iron) due to trapping of interstitials by impurities. Suppression of these defects is observed in MgO:H crystals and attributed to migration of oxygen vacancies to microcavities filled with H 2 gas. In MgO:Li crystals irradiated with neutron fluences below 10 18 n/cm 2 , most of the oxygen vacancies are camouflaged as hydride ions. Nanoindentation experiments show that hardness increases with neutron fluence and is independent of the presence of lithium in the crystal. Comparison between a neutron-irradiated and a thermochemically reduced crystal containing similar concentrations of oxygen vacancies shows that 70% of the neutron-irradiation hardening is produced by interstitials, 30% by oxygen vacancies and a negligible amount by higher-order point defects

  6. Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renault Laborne, Alexandra, E-mail: alexandra.renault@cea.fr [CEA, DEN, SRMA, F-91191 Gif-sur-Yvette (France); Gavoille, Pierre [CEA, DEN, SEMI, F-91191 Gif-sur-Yvette (France); Malaplate, Joël [CEA, DEN, SRMA, F-91191 Gif-sur-Yvette (France); Pokor, Cédric [EDF R& D, MMC, Site des Renardières, F-77818 Morêt-sur-Loing cedex (France); Tanguy, Benoît [CEA, DEN, SEMI, F-91191 Gif-sur-Yvette (France)

    2015-05-15

    Annealed specimens of type 304L and 316 stainless steel and cold-worked 316 specimens were irradiated in the Phénix reactor in the temperature range 381–394 °C and to different damage doses up to 39 dpa. The microstructure and microchemistry of both 304L and 316 have been examined using the combination of the different techniques of TEM to establish the void swelling and precipitation behavior under neutron irradiation. TEM observations are compared with results of measurements of immersion density and thermo-electric power obtained on the same irradiated stainless steels. The similarities and differences in their behavior on different scales are used to understand the factors in terms of the chemical composition and metallurgical state of steels, affecting the precipitation under irradiation and the swelling behavior. Irradiation induces the formation of some precipitate phases (e.g., M{sub 6}C and M{sub 23}C{sub 6}-type carbides, and γ’- and G-phases), Frank loops and cavities. According to the metallurgical state and chemical composition of the steel, the amount of each type of radiation-induced defects is not the same, affecting their density and thermo-electric power.

  7. Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

    Science.gov (United States)

    Renault Laborne, Alexandra; Gavoille, Pierre; Malaplate, Joël; Pokor, Cédric; Tanguy, Benoît

    2015-05-01

    Annealed specimens of type 304L and 316 stainless steel and cold-worked 316 specimens were irradiated in the Phénix reactor in the temperature range 381-394 °C and to different damage doses up to 39 dpa. The microstructure and microchemistry of both 304L and 316 have been examined using the combination of the different techniques of TEM to establish the void swelling and precipitation behavior under neutron irradiation. TEM observations are compared with results of measurements of immersion density and thermo-electric power obtained on the same irradiated stainless steels. The similarities and differences in their behavior on different scales are used to understand the factors in terms of the chemical composition and metallurgical state of steels, affecting the precipitation under irradiation and the swelling behavior. Irradiation induces the formation of some precipitate phases (e.g., M6C and M23C6-type carbides, and γ'- and G-phases), Frank loops and cavities. According to the metallurgical state and chemical composition of the steel, the amount of each type of radiation-induced defects is not the same, affecting their density and thermo-electric power.

  8. Correlation of radiation-induced changes in microstructure/microchemistry, density and thermo-electric power of type 304L and 316 stainless steels irradiated in the Phénix reactor

    International Nuclear Information System (INIS)

    Renault Laborne, Alexandra; Gavoille, Pierre; Malaplate, Joël; Pokor, Cédric; Tanguy, Benoît

    2015-01-01

    Annealed specimens of type 304L and 316 stainless steel and cold-worked 316 specimens were irradiated in the Phénix reactor in the temperature range 381–394 °C and to different damage doses up to 39 dpa. The microstructure and microchemistry of both 304L and 316 have been examined using the combination of the different techniques of TEM to establish the void swelling and precipitation behavior under neutron irradiation. TEM observations are compared with results of measurements of immersion density and thermo-electric power obtained on the same irradiated stainless steels. The similarities and differences in their behavior on different scales are used to understand the factors in terms of the chemical composition and metallurgical state of steels, affecting the precipitation under irradiation and the swelling behavior. Irradiation induces the formation of some precipitate phases (e.g., M 6 C and M 23 C 6 -type carbides, and γ’- and G-phases), Frank loops and cavities. According to the metallurgical state and chemical composition of the steel, the amount of each type of radiation-induced defects is not the same, affecting their density and thermo-electric power

  9. Effect of additional minor elements on accumulation behavior of point defects under electron irradiation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Sekio, Yoshihiro; Yamashita, Shinichiro; Takahashi, Heishichiro; Sakaguchi, Norihito

    2014-01-01

    Addition of minor elements to a base alloy is often applied with the aim of mitigating void swelling by decreasing the vacancy diffusivity and flux which influence vacancy accumulation behavior. However, the comparative evaluations of parameters, such as the diffusivity and flux, between a base alloy and modified alloys with specific additives have not been studied in detail. In this study, type 316 austenitic stainless steel as a base alloy and type 316 austenitic stainless steels modified with vanadium (V) or zirconium (Zr) additions were used to perform evaluations from the changes of widths of the void denuded zone (VDZ) formed near a random grain boundary during electron irradiation because these widths depend on vacancy diffusivity and flux. The formations of VDZs were observed in in-situ observations during electron irradiation at 723 K and the formed VDZ widths were measured from the transmission electron microscopic images after electron irradiation. As a result, the VDZs were formed in both steels without and with V, and respective widths were ∼119 and ∼100 nm. On the other hand, the VDZ formation was not observed clearly in the steel with Zr. From the measured VDZ widths in the steels without and with V addition, the estimated ratio of the vacancy diffusivity in the steel with V to that in the steel without V was about 0.50 and the estimated ratio of the vacancy flux in the steel with V to that in the steel without V was about 0.71. This result suggests that the effect of additional minor elements on vacancy accumulation behaviors under electron irradiation could be estimated from evaluations of the VDZ width changes among steels with and without minor elements. Especially, because void swelling is closely related with the vacancy diffusion process, the VDZ width changes would also be reflected on void swelling behavior. (author)

  10. Using TRIGA Mark II research reactor for irradiation with thermal neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

    2015-03-15

    Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

  11. EPR of alanine irradiated by neutrons

    International Nuclear Information System (INIS)

    Pivovarov, S.P.; Seredavina, T.A.; Zhdanov, S.V.; Mul'gin, S.I.; Zhakparov, R.K.

    2001-01-01

    In the work the first results of EPR studies of alanine, irradiated with diverse doses at neutron cyclotron generator different conditions and on the critical reactor stand are presented. A dose linearity dependence of EPR signal is observing, the methods of γ-background contribution separation are discussed. Obtain results is giving the basis to recommendation of alanine as an effective detector irradiation. However it is demanded the farther study on clarification of radiation sensitivity value dependence on the neutron energy spectrum form

  12. Dissipation and accumulation of energy during plastic deformation of Armco -iron and 12Cr18Ni10Ti stainless steel irradiated by neutrons

    International Nuclear Information System (INIS)

    Toktogulova, D.; Maksimkin, O.; Gusev, M.; Garner, F.

    2007-01-01

    Full text of publication follows: Much attention is currently being paid in the fusion materials community to modeling of radiation damage and its consequences in structural alloys on mechanical properties. Such activities are best guided with experimental data on the fundamental microstructural and thermodynamic processes involved. This report addresses such fundamental concerns. During plastic deformation of metals some fraction of the externally-applied mechanical energy is converted into heat and is partially accumulated in the form of crystal lattice defects. The thermal release arises from gliding dislocations, their various interactions, their annihilation etc. With respect to irradiated material, one might expect additional heat release caused by interactions of dislocation and radiation-induced defects. To explore this possibility flat mini-tensile specimens of Armco-iron and 12Cr18Ni10Ti stainless steel, both in the annealed condition, were irradiated in the range 2x10 18 to 1.3x10 20 n/cm 2 (E>0.1 MeV) in the WWR-K reactor at T≤350 K. Mechanical tests of both irradiated and non-irradiated specimens were conducted at room temperature in a facility that was a combination of a Calvet calorimeter and a micro-tensile device. This allows simultaneous measurement of mechanical properties and thermodynamic parameters such as deformation work, dissipated heat and latent energy during deformation. The authors derived the kinetics of changes in thermodynamic characteristics versus the deformation level. As the neutron fluence rises, the material's capability to accumulate energy appears to be declining. For example, 12Cr18Ni10Ti irradiated to 1.3x10 20 n/cm 2 did not show any energy accumulation under deformation. In Armco-iron at 1.4x10 19 n/cm 2 the heat release considerably exceeded the deformation work value. The authors assume that such effects might be related with annihilation of point defects and their complexes introduced during irradiation. To test this

  13. 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

  14. Defect recovery in proton irradiated Ti-modified stainless steel probed by positron annihilation

    Energy Technology Data Exchange (ETDEWEB)

    Arunkumar, J.; Abhaya, S.; Rajaraman, R.; Amarendra, G. [Materials Science Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu 603102 (India); Nair, K.G.M. [Materials Science Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu 603102 (India)], E-mail: kgmn@igcar.gov.in; Sundar, C.S.; Raj, Baldev [Materials Science Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu 603102 (India)

    2009-02-28

    The defect recovery in proton irradiated Ti-modified D9 steel has been studied by positron annihilation isochronal and isothermal annealing measurements. D9 samples have been irradiated with 3 MeV protons followed by isochronal annealing at various temperatures in the range of 323 to 1273 K. The dramatic decrease in positron annihilation parameters, viz. positron lifetime and Doppler S-parameter, around 500 K indicates the recovery of vacancy-defects. A clear difference in the recovery beyond 700 K is observed between solution annealed and cold worked state of D9 steel due to the precipitation of TiC in the latter. Isothermal annealing studies have been carried out at the temperature wherein vacancies distinctly migrate. Assuming a singly activated process for defect annealing, the effective activation energy for vacancy migration is estimated to be 1.13 {+-} 0.08 eV.

  15. Observation of oscillatory radiation induced segregation profiles at grain boundaries in neutron irradiated 316 stainless steel using atom probe tomography

    Science.gov (United States)

    Barr, Christopher M.; Felfer, Peter J.; Cole, James I.; Taheri, Mitra L.

    2018-06-01

    Radiation induced segregation in austenitic Fe-Ni-Cr stainless steels is a key detrimental microstructural modification experienced in the current generation of light water reactors. In particular, Cr depletion at grain boundaries can be a significant factor in irradiation-assisted stress corrosion cracking. Therefore, having a complete knowledge and mechanistic understanding of radiation induced segregation at high dose and after a long thermal history is desired for continued sustainability of existing reactors. Here, we examine a 12% cold worked AISI 316 stainless steel hexagonal duct exposed in the lower dose, outer blanket region of the EBR-II reactor, by using advanced characterization and analysis techniques including atom probe tomography and analytical scanning transmission electron microscopy. Contrary to existing literature, we observe an oscillatory w-shape Cr and M-shape Ni concentration profile at 31 dpa. The presence and characterization through advanced atom probe tomography analysis of the w-shape Cr RIS profile is discussed in the context of the localized GB plane interfacial excess of the other major and minor alloying elements. The key finding of a co-segregation phenomena coupling Cr, Mo, and C is discussed in the context of the existing solute segregation literature under irradiation with emphasis on improved spatial and chemical resolution of atom probe tomography.

  16. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

    International Nuclear Information System (INIS)

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-01-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1x10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6 C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6 C in JPCA-2. (orig.)

  17. F-type centers in neutron-irradiated AIN

    International Nuclear Information System (INIS)

    Atobe, Kozo; Honda, Makoto; Fukuoka, Noboru; Okada, Moritami; Nakagawa, Masuo.

    1990-01-01

    The production of point defects by neutron irradiation and thermal decay in sintered AIN polycrystal are investigated. The absorption band at 370 nm is observed after reactor neutron irradiation to a dose of 10 16 n/cm 2 (E > 0.1 MeV). The defect corresponding to the band is tentatively assigned as an F-type center from the optical absorption and electron spin resonance. (author)

  18. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2016-01-22

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  19. Effects of irradiation and thermal aging upon fatigue-crack growth behavior of reactor pressure boundary materials. [Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    James, L. A.

    1978-10-01

    Two processes that have the potential to produce degradation in the properties of pressure boundary materials are neutron irradiation and long-time thermal aging. This paper uses linear-elastic fracture mechanics techniques to assess the effect of these two processes upon the fatigue-crack growth behavior of a number of alloys commonly employed in reactor pressure boundaries. The materials evaluated include ferritic steels, austenitic stainless steels, and nickel-base alloys typical of those employed in a number of reactor types including water-cooled, gas-cooled, and liquid-metal-cooled designs.

  20. Neutron and gamma irradiation damage to organic materials.

    Energy Technology Data Exchange (ETDEWEB)

    White, Gregory Von, II; Bernstein, Robert

    2012-04-01

    This document discusses open literature reports which investigate the damage effects of neutron and gamma irradiation on polymers and/or epoxies - damage refers to reduced physical chemical, and electrical properties. Based on the literature, correlations are made for an SNL developed epoxy (Epon 828-1031/DDS) with an expected total fast-neutron fluence of {approx}10{sup 12} n/cm{sup 2} and a {gamma} dosage of {approx}500 Gy received over {approx}30 years at < 200 C. In short, there are no gamma and neutron irradiation concerns for Epon 828-1031/DDS. To enhance the fidelity of our hypotheses, in regards to radiation damage, we propose future work consisting of simultaneous thermal/irradiation (neutron and gamma) experiments that will help elucidate any damage concerns at these specified environmental conditions.

  1. Evolution of secondary-phase precipitates during annealing of the 12Kh18N9T steel irradiated with neutrons to a dose of 5 DPA

    Science.gov (United States)

    Tsai, K. V.; Maksimkin, O. P.; Turubarova, L. G.

    2007-03-01

    The formation and evolution of thermally-induced secondary precipitates in an austenitic stainless steel 12Kh18N9T irradiated in the core of a laboratory reactor VVR-K to a dose of 5 dpa and subjected to post-radiation isochronous annealings for 1 h in a temperature range from 450 to 1050°C have been studied using transmission electron microscopy (TEM) and microhardness measurements. It has been shown that the formation of stitch (secondary) titanium carbides and M 23C6 carbides at grain and twin boundaries after annealing at 1050°C is preceded by a complex evolution of fineparticles of secondary phases (titanium carbides and nitrides) precipitated at dislocation loops and dislocations during annealing at temperatures above 750°C.

  2. Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

    International Nuclear Information System (INIS)

    Little, E.A.; Stoter, L.P.

    1982-01-01

    The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

  3. Influence of neutron irradiation on ferromagnetic metallic glasses

    International Nuclear Information System (INIS)

    Miglierini, M.; Nasu, Saburo; Sitek, J.

    1992-01-01

    Transmission 57 Fe Moessbauer spectroscopy is used to study effects of neutron irradiation on magnetic properties of Fe-based ferromagnetic metallic glasses. Elastic stress centers are produced during the process of neutron irradiation as a result of atom mixing. Rearrangement of the atoms causes changes in the average value of the hyperfine field distribution and orientation of the net magnetic moment. They are shown to depend on the composition of the investigated samples. Cr-doped metallic glasses depict transformation from ferromagnetic to paramagnetic state at room temperature after neutron irradiation implying changes in the Curie temperature. Presence of Ni in the samples reduces the effects of radiation damage. (orig.)

  4. Neutron Focusing Mirrors for Neutron Radiography of Irradiated Nuclear Fuel at Idaho National Laboratory

    Science.gov (United States)

    Rai, Durgesh K.; Wu, Huarui; Abir, Muhammad; Giglio, Jeffrey; Khaykovich, Boris

    Post irradiation examination (PIE) of samples irradiated in nuclear reactors is a challenging but necessary task for the development on novel nuclear power reactors. Idaho National Laboratory (INL) has neutron radiography capabilities, which are especially useful for the PIE of irradiated nuclear fuel. These capabilities are limited due to the extremely high gamma-ray radiation from the irradiated fuel, which precludes the use of standard digital detectors, in turn limiting the ability to do tomography and driving the cost of the measurements. In addition, the small 250 kW Neutron Radiography Reactor (NRAD) provides a relatively weak neutron flux, which leads to low signal-to-noise ratio. In this work, we develop neutron focusing optics suitable for the installation at NRAD. The optics would separate the sample and the detector, potentially allowing for the use of digital radiography detectors, and would provide significant intensity enhancement as well. The optics consist of several coaxial nested Wolter mirrors and is suited for polychromatic thermal neutron radiation. Laboratory Directed Research and Development program of Idaho National Laboratory.

  5. Use of microhardness to determine the strengthening and microstructural alterations of 14-MeV-neutron-irradiated metals

    International Nuclear Information System (INIS)

    Panayotou, N.F.

    1982-02-01

    Microhardness has been found to be an effective post-irradiation analytical tool. The hardening of recrystallized copper and type 316 stainless steel irradiated up to 1 x 10 18 n/cm 2 , E = 14 MeV, at the Rotating Target Neutron Source (RTNS)-II has been studied. It was determined that for these metals the increase in hardness varies directly with the measured increase in the 0.2 percent offset yield strength. Furthermore the coupled use of microhardness and transmission electron microscope techniques provides an estimate of the defect population which was not resolved by TEM. This estimate in turn, was used to evaluate the magnitude of the proportionality constant used in the strong barrier model of radiation induced hardening

  6. Stress corrosion cracking of irradiated stainless steels in primary water: experimental studies and model development in the FP7 PERFECT IP

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2009-01-01

    The long term behaviour of the internal structures, surrounding the core of nuclear reactors, is a concern within the general framework of plant life management. Due to their positioning in the reactor, the internal structures of a pressurised water reactor (PWR) receive a high neutron irradiation dose during their exposure to the primary environment. The irradiation induced changes in the material and environment may cause or accelerate stress corrosion cracking of stainless steel internals, otherwise insensitive to stress corrosion in primary environment. This phenomenon of irradiation assisted stress corrosion cracking (IASCC) is currently not well quantified. At this stage, the understanding of the underlying mechanisms of IASCC is qualitative at best and no systematic database and descriptive model are available for IASCC in PWR conditions. Since 2004, a concerted European effort is ongoing within the framework 7 PERFECT integrated project to investigate and model IASCC

  7. Neutron Absorbing Ability Variation in Neutron Absorbing Material Caused by the Neutron Irradiation in Spent Fuel Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Hee Dong; Han, Seul Gi; Lee, Sang Dong; Kim, Ki Hong; Ryu, Eag Hyang; Park, Hwa Gyu [Doosan Heavy Industries and Construction, Changwon (Korea, Republic of)

    2014-10-15

    In spent fuel storage facility like high density spent fuel storage racks and dry storage casks, spent fuels are stored with neutron absorbing materials installed as a part of those facilities, and they are used for absorbing neutrons emitted from spent fuels. Usually structural material with neutron absorbing material of racks and casks are located around spent fuels, so it is irradiated by neutrons for long time. Neutron absorbing ability could be changed by the variation of nuclide composition in neutron absorbing material caused by the irradiation of neutrons. So, neutron absorbing materials are continuously faced with spent fuels with boric acid solution or inert gas environment. Major nuclides in neutron absorbing material are Al{sup 27}, C{sup 12}, B{sup 11}, B{sup 10} and they are changed to numerous other ones as radioactive decay or neutron absorption reaction. The B{sup 10} content in neutron absorbing material dominates the neutron absorbing ability, so, the variation of nuclide composition including the decrease of B{sup 10} content is the critical factor on neutron absorbing ability. In this study, neutron flux in spent fuel, the activation of neutron absorbing material and the variation of nuclide composition are calculated. And, the minimum neutron flux causing the decrease of B{sup 10} content is calculated in spent fuel storage facility. Finally, the variation of neutron multiplication factor is identified according to the one of B{sup 10} content in neutron absorbing material. The minimum neutron flux to impact the neutron absorbing ability is 10{sup 10} order, however, usual neutron flux from spent fuel is 10{sup 8} order. Therefore, even though neutron absorbing material is irradiated for over 40 years, B{sup 10} content is little decreased, so, initial neutron absorbing ability could be kept continuously.

  8. 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.)

  9. Titanium dioxide modified with various amines used as sorbents of carbon dioxide

    International Nuclear Information System (INIS)

    Kapica-Kozar, Joanna; Pirog, Ewa; Kusiak-Nejman, Ewelina; Wrobel, Rafal J.; Gesikiewicz-Puchalska, Andzelika; Morawski, Antoni W.; Narkiewicz, Urszula; Michalkiewicz, Beata

    2017-01-01

    In this study, titanium dioxide was modified with various amines through hydrothermal treatment for adsorption of CO_2. The carbon dioxide adsorption performance of the prepared samples was measured using an STA 449 C thermo-balance (Netzsch Company, Germany). The morphological structures, functional groups and elemental compositions of the unmodified and amine-modified titanium dioxide sorbents were characterized by X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FT-IR/DR) and scanning electron microscopy (SEM), respectively. The results showed that modification of TiO_2 with amines through hydrothermal treatment is a simple method to prepare CO_2 sorbents with high adsorption capacities. Moreover, the results revealed that TEPA-modified titanium dioxide shoved the highest adsorption capacity, enabling an increase in CO_2 uptake from 0.45 mmol CO_2 g"-"1 in the case of raw TiO_2 to 1.63 mmol CO_2 g"-"1. This result could be indirectly related to the fact that TEPA has the highest amino group content among the three amines used in our research. Additionally, durability tests performed by cyclic adsorption-desorption revealed that TEPA modified titanium dioxide also possesses excellent stability, despite a slight decrease in adsorption capacity over time. (authors)

  10. A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

    A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

  11. Adsorbed radioactivity and radiographic imaging of surfaces of stainless steel and titanium

    Science.gov (United States)

    Jung, Haijo

    1997-11-01

    Type 304 stainless steel used for typical surface materials of spent fuel shipping casks and titanium were exposed in the spent fuel storage pool of a typical PWR power plant. Adsorption characteristics, effectiveness of decontamination by water cleaning and by electrocleaning, and swipe effectiveness on the metal surfaces were studied. A variety of environmental conditions had been manipulated to stimulate the potential 'weeping' phenomenon that often occurs with spent fuel shipping casks during transit. In a previous study, few heterogeneous effects of adsorbed contamination onto metal surfaces were observed. Radiographic images of cask surfaces were made in this study and showed clearly heterogeneous activity distributions. Acquired radiographic images were digitized and further analyzed with an image analysis computer package and compared to calibrated images by using standard sources. The measurements of activity distribution by using the radiographic image method were consistent with that using a HPGe detector. This radiographic image method was used to study the effects of electrocleaning for total and specified areas. The Modulation Transfer Function (MTF) of a film-screen system in contact with a radioactive metal surface was studied with neutron activated gold foils and showed more broad resolution properties than general diagnostic x-ray film-screen systems. Microstructure between normal areas and hot spots showed significant differences, and one hot spot appearing as a dot on the film image consisted of several small hot spots (about 10 μm in diameter). These hot spots were observed as structural defects of the metal surfaces.

  12. Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

    The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

  13. Modification of the surfaces of stainless steel during titanium nitride deposition by a dynamic mixing method

    Science.gov (United States)

    Yokota, Katsuhiro; Tamura, Susumu; Nakamura, Kazuhiro; Horiguchi, Motohiro; Nakaiwa, Hiroki; Sugimoto, Takashi; Akamatsu, Katsuya; Nakao, Kazuyoshi

    2000-05-01

    Surfaces of stainless steel SUS304 were coated with titanium nitride (TiN) at temperatures ranging from 400°C to 770°C using a dynamic mixing technique. The N+ ions were accelerated at energies of 0.5-2.0 keV, and were implanted into the stainless steel. The composition of the prepared TiN films was measured using Rutherford backscattering spectrometry with He ions at an energy of 2.0 MeV. Intermediate layers containing compounds such as FesNq, Cr2N, and CrFe were formed between the TiN films and substrates at substrate temperatures higher than 700°C. The thickness of the TiN films decreased significantly when the intermediate layers were formed.

  14. Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

    2016-10-15

    Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

  15. Radiation-thermal effects change of physico-mechanical properties in reactor materials irradiated with neutrons and energetic charged particles

    International Nuclear Information System (INIS)

    Hofman, A.

    1999-01-01

    In the first part of the report (chapter 1) the earlier results of the important scientific and technological investigations which were performed in the seventies years in Poland have been presented. They concerned the fabrication, corrosion, mechanical properties of materials for research and power reactors. Being of the general survey character, the chapter includes own, original results of research of thermal irradiation effects on microstructure evolution phase transformations and mechanical properties of reactor materials. The kinetics of isothermal transformation β→α in U-Cr 0.4% wt. alloy has been studied. Factors affecting stress-corrosion cracking of zirconium in iodine vapour have been investigated. The rings and loops for irradiation specimens and Hot Laboratory for postirradiation examination of construction materials is described. In the second part (chapters 2, 3, 4, 5) performed the investigations and simulations of radiation damage in metals by heavy ion beams (E > 1 MeV/a.m.n.) were described scientific base and technical problems of the method of irradiation of heavy ions and of the examination of irradiated samples is presented. It is followed by a summary of the results of simulation and reactor experiments on different materials. Radiation hardening of a number metals (Al, Zr, Cu, Ni, U) irradiated by heavy ion and neutrons, mechanical properties and microstructural evolution in ion and neutron irradiated austenitic stainless steel is described. The last chapter is a description of practical aspects of the presented studies in nuclear science and technology. (author)

  16. Coating of Titanium Nitride on Stainless Steel Targets by a 4 kJ Plasma Focus Device

    Science.gov (United States)

    Omrani, M.; Habibi, M.; Amrollahi, R.

    2012-08-01

    Titanium nitride thin films were deposited on stainless steel (SS316L) targets by using a 4 kJ plasma focus device. The corresponding energy flux delivered to SS316L surface is estimated to be 2.69 × 1013 kev cm-3 ns-1. X-ray diffraction analysis reveals the formation of a nanocrystalline titanium nitride coating on the surface of targets. Thickness of the elements found on the surface of treated samples which are obtained by Rutherford backscattering spectrometry analysis (RBS) were (×1015 at/cm2) .45% Ti, 50% N and 5% Fe. Scanning electron microscopy was used to indicate changes in surface morphology. Existence of grains in different size confirms the formation of TiN crystals on the surface of targets.

  17. Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging*1

    Science.gov (United States)

    Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

    1988-07-01

    Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6C in JPCA-2.

  18. Proceedings of neutron irradiation technical meeting on BNCT

    International Nuclear Information System (INIS)

    2000-10-01

    The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

  19. Feasibility of long-life and corrosion-resistant canister with titanium cladding

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Tokiwai, Moriyasu; Saegusa, Toshiari

    2008-01-01

    In order to store nuclear spent fuels for a long term, we propose the concept of stainless steel canister with titanium cladding. The stainless canister is first brazed to titanium plates, and then the brazed joints are covered with other titanium plates. A MIG brazing for titanium and stainless steel was demonstrated with a brazing metal of Cu-1Mn-3Si alloy (MG960). JIS G 0601 shear strength, tensile shear stress and peel strength tests are conducted for the optimized MIG brazing conditions. These results showed the MIG brazing specimens possess adequate structural strength. After the salt spray test on the basis of JIS Z 2371, there were no pitting and general corrosions on a TIG welding specimen between titanium plates. The corrosion resistance is therefore, sufficiently high. Manufacturing cost estimation suggests that the titanium cladding concept is feasible thereby using 1-mm-thick titanium plates to reduce the material cost. In addition to this concept, we propose another concept of the canister by using titanium-stainless steel cladding plates to reduce a number of brazing joints. (author)

  20. Evidence of emission of neutrons from a titanium-deuterium system

    International Nuclear Information System (INIS)

    Ninno, A. de; Frattolillo, A.; Lollobattista, G.; Martinis, L.; Martone, M.; Mori, L.; Podda, S.; Scaramuzzi, F.

    1989-01-01

    The interaction of deuterium gas with titanium has produced a flow of neutrons in two experiments reported here. This seems to show that it is not necessary to use electrolysis in order to obtain a low-temperature fusion reaction between deuterium nuclei. The experiment confirms also that nonequilibrium conditions are necessary in order to produce such a phenomenon

  1. Micro-Raman and photoluminescence studies of neutron-irradiated gallium nitride epilayers

    International Nuclear Information System (INIS)

    Wang, R.X.; Xu, S.J.; Fung, S.; Beling, C.D.; Wang, K.; Li, S.; Wei, Z.F.; Zhou, T.J.; Zhang, J.D.; Huang Ying; Gong, M.

    2005-01-01

    GaN epilayers grown on sapphire substrate were irradiated with various dosages of neutrons and were characterized using Micro-Raman and photoluminescence. It was found that the A 1 (LO) peak in the Raman spectra clearly shifted with neutron irradiation dosage. Careful curve fitting of the Raman data was carried out to obtain the carrier concentration which was found to vary with the neutron irradiation dosage. The variation of the full width at half maximum height of the photoluminescence was consistent with the Raman results. The neutron irradiation-induced structural defects (likely to be Ge Ga ) give rise to carrier trap centers which are responsible for the observed reduction in carrier concentration of the irradiated GaN

  2. Influence of titanium on the tempering structure of austenitic steels

    International Nuclear Information System (INIS)

    Ghuezaiel, M.J.

    1985-10-01

    The microstructure of titanium-stabilized and initially deformed (approximately 20%) austenitic stainless steels used in structures of fast neutrons reactors has been studied after one hour duration annealings (500 0 C) by X-ray diffraction, optical microscopy, microhardness and transmission electron microscopy. The studied alloys were either of industrial type CND 17-13 (0.23 to 0.45 wt% Ti) or pure steels (18% Cr, 14% Ni, 0 or 0.3 wt% Ti). During tempering, the pure steels presented some restauration before recristallization. In the industrial steels, only recristallization occurred, and this only in the most deformed steel. Precipitation does not occur in the titanium-free pure steel. In industrial steels, many intermetallic phases are formed when recristallization starts [fr

  3. A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

    1999-01-01

    In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

  4. Irradiation temperature dependence of defect formation of nitrides (A1N and c-BN) during neutron irradiations

    International Nuclear Information System (INIS)

    Atobe, Kozo.; Okada, Moritami; Nakagawa, Masuo

    2000-01-01

    The nitrogen vacancy concentration in the more refractory nitrides (A1N and c-BN) is determined as a function of reactor fluence up to 5.2x10 17 thermal neutrons/cm 2 and a function of the irradiation temperature at 25, 50, 100, 150, 200, 250 K. It is found that there is no remarkable dependence of the defect formation in nitrides on the irradiation temperature. The production of damage in the nitrides is considerably different from that in oxides. From the irradiation experiments using thermal neutron irradiation field, it is suggested in reactor irradiation that the atomic displacements in the nitrides occur predominately from energetic particles of the nuclear reactions with thermal neutrons in addition to the elastic collisions by fast neutron

  5. Positron lifetime study of copper irradiated by energetic protons or energetic neutrons

    International Nuclear Information System (INIS)

    Howell, R.H.

    1979-03-01

    Positron lifetime measurements of pure copper damaged by irradiation with energetic protons and neutrons are presented. Lifetime determinations of the bulk material and various traps were made, and the dependence of the trapping rate on dose and irradiation energy were investigated. The results from the neutron- and proton-irradiated samples point to the existence of traps with similar but distinct lifetime parameters, not varying greatly from values reported in deformation studies. Also, a trap with long lifetime is seen for some proton irradiations, but is never seen for the neutron irradiations. The trapping rate of the short-lifetime trap is a linear function of dose for proton-irradiated samples and nearly so for the neutron irradiation. 1 figure

  6. Microstructural stability of austenitic stainless steels on exposure to irradiation and elevated temperatures

    International Nuclear Information System (INIS)

    Parameswaran, P.; Radhika, M.; Saroja, S.; Vijayalakshmi, M.; Nanda Gopal, M.

    2011-01-01

    Cold worked 316 stainless steels employed as core material in fast reactors on exposure to neutron irradiation to 40 dpa at ∼ 450 deg C have resulted in microstructural changes in terms of formation of voids and extensive precipitation of carbides, eta phase and nickel silicides. As a consequence there is degradation in the mechanical properties of the material, particularly ductility. In order to achieve higher burnup it is essential to find better materials, which would exhibit less void swelling and retain the microstructure over long radiation doses. Accordingly alloy D9 with appropriate modifications of Ni and Cr content with Ti additions has been developed. Further modification of alloy D9 with respect to minor alloying additions namely Si and P is being studied, in order to enhance the radiation resistance for extending the service life of components. The effectiveness of these elements can be achieved if and only if they are retained in solution over long time of exposure at high temperatures and irradiation. Therefore, the thermal stability of the newly developed improved D9 alloys, with a constant Ti:C ratio and different levels of Si and P has been studied with respect to microstructural evolution and its influence on the mechanical properties. Thermal aging behavior of the alloy with varying titanium contents at elevated temperatures was also studied in detail to identify the optimum alloying levels. The alloys in the 20% cold worked condition exhibit austenitic grains interspersed with bands of fine cold worked grains. On aging in the temperature range of 873-1073K for various durations upto two years the alloy showed the presence of different phases such as M 23 C 6 , intermetallics and TiC whose quantity varies with temperature. The hardness values showed a trend of an initial increase in all the alloys but at longer times the hardness either showed saturation or a decrease followed by saturation. The microstructural parameters like grain size and

  7. 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

  8. TEM examination of the effect of post-irradiation annealing on 7.7 dpa AISI 304 stainless steel

    International Nuclear Information System (INIS)

    Karlsen, W.; Ivanchenko, M.; Pakarinen, J.; Karlsen, T.

    2015-01-01

    Stainless steels exposed to neutron irradiation during service in light water reactors (LWR) can become susceptible to intergranular cracking, referred to as irradiation assisted stress corrosion cracking (IASCC). Analytical transmission electron microscopy (ATEM) was used to examine the effect of post-irradiation annealing (PIA) on radiation-induced segregation (RIS) at the grain boundaries of 7.7 dpa AISI 304 stainless steel. The grain boundary profiles and the irradiation damage were analysed in the as-irradiated state and after PIA of 6 hours at 500 C. degrees and after 25 hours at 500 C. degrees and 550 C. degrees by using transmission electron microscopy (TEM). As a main conclusion from the TEM examinations, the effects of PIA were found to be relatively small after only 6 hours, while after 25 hours of PIA at both 500 and 550 C. degrees, RIS was almost recovered and only marginal deviation in chemical composition could be found near the GB. The as-irradiated state showed extreme RIS values of Si 4.9 wt%, Cr 14.7 wt%, Ni 23.4 wt%, and P 1.4 wt%., while upon PIA for 6 hours the extreme values for RIS were Si 3.9 wt%, Cr 16.0 wt%, Ni 21 wt%, and P 0.9 wt%. After 6 hours annealing at 500 C. dislocation loops start to grow, while dislocation density remains of the same order of magnitude. After annealing for 25 hours at 500 C. degrees the average size of dislocation loops remains nearly the same, while dislocation density was reduced almost by one fold. In the areas where dislocation density was found to be the lowest some features, which can most likely be attributed to stacking fault tetrahedral (SFT) were found. Annealing at even higher temperature (550 C.) affected the average size of the dislocation loops, making them almost twice as large as well as resulting in a very broad distribution of dislocation sizes. Density of dislocations is also reduced by one fold in comparison to the as irradiated condition and leads to formation of SFTs, which could be

  9. Does surface anodisation of titanium implants change osseointegration and make their extraction from bone any easier?

    OpenAIRE

    Langhoff, J; Mayer, J; Faber, L; Kästner, S B; Guibert, G; Zlinszky, K; Auer, J A; von Rechenberg, B

    2008-01-01

    Objectives: Titanium implants have a tendency for high bone-implant bonding, and, in comparison to stainless steel implants are more difficult to remove. The current study was carried out to evaluate, i) the release strength of three selected anodized titanium surfaces with increased nanohardness and low roughness, and ii) bone-implant bonding in vivo. These modified surfaces were intended to give improved anchorage while facilitating easier removal of temporary implants. Material and methods...

  10. Instrumental neutron-activation determination of impurities in lead and titanium compounds

    Energy Technology Data Exchange (ETDEWEB)

    Popova, I L

    1980-01-01

    Instrumental neutron-activation analysis was used to determine 22 impurities in lead and titanium compounds (e.g. PbO, Pb/NO3/2, and TiO2) used as raw materials for ferroelectrics. Five elements (Al, V, Mn, Sc, and Se) were determined by short-lived isotopes and 17 elements were determined by long-lived isotopes. The detection limits were 7 x 10 to the -3rd to 2 x 10 to the -8th %. A substantial difference in concentrations of certain impurity elements has been found in different series of lead and titanium oxides of similar purity.

  11. Microstructure and mechanical properties of titanium aluminum carbides neutron irradiated at 400–700 °C

    Energy Technology Data Exchange (ETDEWEB)

    Ang, Caen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Parish, Chad M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Shih, Chunghao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); General Atomics, San Diego, CA (United States); Silva, Chinthaka [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-02-23

    Here, this work reports the first mechanical properties of Ti3AlC2-Ti5Al2C3 materials neutron irradiated at ~400, 630 and 700 °C at a fluence of 2 × 1025 n m-2 (E > 0.1 MeV) or a displacement dose of ~2 dpa. After irradiation at ~400 °C, anisotropic swelling and loss of 90% flexural strength was observed. After irradiation at ~630–700 °C, properties were unchanged. Microcracking and kinking-delamination had occurred during irradiation at ~630–700 °C. Further examination showed no cavities in Ti3AlC2 after irradiation at ~630 °C, and MX and A lamellae were preserved. However, disturbance of (0004) reflections corresponding to M-A layers was observed, and the number density of line/planar defects was ~1023 m-3 of size 5–10 nm. HAADF identified these defects as antisite TiAl atoms. Finally, Ti3AlC2-Ti5Al2C3 shows abrupt dynamic recovery of A-layers from ~630 °C, but a higher temperature appears necessary for full recovery.

  12. Measurement and modeling of radiation-induced grain boundary grain boundary segregation in stainless steels

    International Nuclear Information System (INIS)

    Bruemmer, S.M.; Charlot, L.A.; Simonen, E.P.

    1995-08-01

    Grain boundary radiation-induced segregation (RIS) in Fe-Ni-Cr stainless alloys has been measured and modelled as a function of irradiation temperature and dose. Heavy-ion irradiation was used to produce damage levels from 1 to 20 displacements per atom (dpa) at temperatures from 175 to 550 degrees C. Measured Fe, Ni, and Cr segregation increased sharply with irradiation dose (from 0 to 5 dpa) and temperature (from 175 to about 350 degrees C). However, grain boundary concentrations did not change significantly as dose or temperatures were further increased. Impurity segregation (Si and P) was also measured, but only Si enrichment appeared to be radiation-induced. Grain boundary Si levels peaked at an intermediate temperature of ∼325 degrees C reaching levels of ∼8 at. %. Equilibrium segregation of P was measured in the high-P alloys, but interfacial concentration did not increase with irradiation exposure. Examination of reported RIS in neutron-irradiated stainless steels revealed similar effects of irradiation dose on grain boundary compositional changes for both major alloying and impurity element's. The Inverse Kirkendall model accurately predicted major alloying element RIS in ion- and neutron-irradiated alloys over the wide range of temperature and dose conditions. In addition, preliminary calculations indicate that the Johnson-Lam model can reasonably estimate grain boundary Si enrichment if back diffusion is enhanced

  13. Thermal analysis of titanium drive-in target for D-D neutron generation.

    Science.gov (United States)

    Jung, N S; Kim, I J; Kim, S J; Choi, H D

    2010-01-01

    Thermal analysis was performed for a titanium drive-in target of a D-D neutron generator. Computational fluid dynamics code CFX-5 was used in this study. To define the heat flux term for the thermal analysis, beam current profile was measured. Temperature of the target was calculated at some of the operating conditions. The cooling performance of the target was evaluated by means of the comparison of the calculated maximum target temperature and the critical temperature of titanium. Copyright 2009 Elsevier Ltd. All rights reserved.

  14. 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)

  15. Bioactivity and osteointegration of hydroxyapatite-coated stainless steel and titanium wires used for intramedullary osteosynthesis.

    Science.gov (United States)

    Popkov, Arnold V; Gorbach, Elena N; Kononovich, Natalia A; Popkov, Dmitry A; Tverdokhlebov, Sergey I; Shesterikov, Evgeniy V

    2017-08-01

    A lot of research was conducted on the use of various biomaterials in orthopedic surgery. Our study investigated the effects of nanostructured calcium-phosphate coating on metallic implants introduced into the bone marrow canal. Stainless steel or titanium 2-mm wires (groups 1 and 2, respectively), and hydroxyapatite-coated stainless steel or titanium wires of the same diameter (groups 3 and 4, respectively) were introduced into the tibial bone marrow canal of 20 dogs (each group = 5 dogs). Hydroxyapatite coating was deposited on the wires with the method of microarc oxidation. Light microscopy to study histological diaphyseal transverse sections, scanning electron microscopy to study the bone marrow area around the implant and an X-ray electron probe analyzer to study the content of calcium and phosphorus were used to investigate bioactivity and osteointegration after a four weeks period. Osteointegration was also assessed by measuring wires' pull-off strength with a sensor dynamometer. Bone formation was observed round the wires in the bone marrow canal in all the groups. Its intensity depended upon the features of wire surfaces and implant materials. Maximum percentage volume of trabecular bone was present in the bone marrow canals of group 4 dogs that corresponded to a mean of 27.1 ± 0.14%, while it was only 6.7% in group 1. The coating in groups 3 and 4 provided better bioactivity and osteointegration. Hydroxyapatite-coated titanium wires showed the highest degree of bone formation around them and greater pull-off strength. Nanostructured hydroxyapatite coating of metallic wires induces an expressed bone formation and provides osteointegration. Hydroxyapatite-coated wires could be used along with external fixation for bone repair enhancement in diaphyseal fractures, management of osteogenesis imperfecta and correction of bone deformities in phosphate diabetes.

  16. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  17. Irradiation creep in reactor graphites for HTR applications. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Veringa, H J; Blackstone, R [Stichting Reactor Centrum Nederland, Petten

    1976-01-01

    A series of restrained shrinkage experiments on a number of graphites in the temperature range 400 to 1400/sup 0/C is described. A description is given of the experimental method and method of data evaluation. The results are compared with data from other sources. Analysis of data confirms that the creep coefficient, which is defined as the radiation induced creep strain per unit stress per unit neutron fluence, is inversely proportional to the pre-irradiation value of the Young's modulus of the material. The radiation creep coefficient increases with temperature in the range 400 to 1400/sup 0/C. It can be represented by the sum of two temperature dependent functions, one of which is inversely proportional to the neutron flux density, the other independent of the neutron flux density. When the data are analysed in this way it is found that the graphites investigated in the present work, although made from widely different starting materials and by different processes, show the same dependence of the irradiation creep coefficient on the temperature and the neutron flux density.

  18. Proceedings of neutron irradiation technical meeting on BNCT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

  19. Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

    1997-10-01

    In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

  20. Effect of Titanium on the Microstructure and Mechanical Properties of High-Carbon Martensitic Stainless Steel 8Cr13MoV

    OpenAIRE

    Wen-Tao Yu; Jing Li; Cheng-Bin Shi; Qin-Tian Zhu

    2016-01-01

    The effect of titanium on the carbides and mechanical properties of martensitic stainless steel 8Cr13MoV was studied. The results showed that TiCs not only acted as nucleation sites for δ-Fe and eutectic carbides, leading to the refinement of the microstructure, but also inhibited the formation of eutectic carbides M7C3. The addition of titanium in steel also promoted the transformation of M7C3-type to M23C6-type carbides, and consequently more carbides could be dissolved into the matrix duri...

  1. Focusing mirrors for enhanced neutron radiography with thermal neutrons and application for irradiated nuclear fuel

    Science.gov (United States)

    Rai, Durgesh K.; Abir, Muhammad; Wu, Huarui; Khaykovich, Boris; Moncton, David E.

    2018-01-01

    Neutron radiography is a powerful method of probing the structure of materials based on attenuation of neutrons. This method is most suitable for materials containing heavy metals, which are not transparent to X-rays, for example irradiated nuclear fuel and other nuclear materials. Neutron radiography is one of the first non-distractive post-irradiated examination methods, which is applied to gain an overview of the integrity of irradiated nuclear fuel and other nuclear materials. However, very powerful gamma radiation emitted by the samples is damaging to the electronics of digital imaging detectors and has so far precluded the use of modern detectors. Here we describe a design of a neutron microscope based on focusing mirrors suitable for thermal neutrons. As in optical microscopes, the sample is separated from the detector, decreasing the effect of gamma radiation. In addition, the application of mirrors would result in a thirty-fold gain in flux and a resolution of better than 40 μm for a field-of-view of about 2.5 cm. Such a thermal neutron microscope can be useful for other applications of neutron radiography, where thermal neutrons are advantageous.

  2. Method of measuring neutron spectra in JMTR exclusively used for irradiation and their evaluation

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    1983-01-01

    In the core of the Japan Materials Testing Reactor, about 60 capsules are irradiated. These are the material capsules for irradiating reactor materials, the fuel capsules for irradiating reactor fuel, the RI capsules for producing radioisotopes and so on. In the irradiation experiment using a reactor, the information on the neutron fluence is indispensable, and the neutron fluence in the irradiated specimen part is evaluated with a dosimeter or the nuclear calculation for the core of the JMTR. At the time of irradiating reactor materials, the dosimeter Fe-54 (n,p) Mn-54 is generally used for evaluating the neutron fluence more than 1 MeV. In the case of fuel irradiation, the thermal neutron fluence is evaluated with the dosimeter Co-59 (n,γ) Co-60. It is important to examine in detail neutron spectra by both calculation and experiment in the reactors exclusively used for irradiation such as the JMTR. The neutron irradiation field in the JMTR, neutron spectrum measuring experiment, the neutron flux monitors for standardizing data, the measurement of X-ray and gamma ray, neutron guess spectrum, the compilation of neutron cross section for SAND 2, and the unfolding of neutron spectra are reported. The degree of agreement of the neutron fluence more than 1 MeV by measurement and calculation was +- 10 to 20 %. (Kako, I.)

  3. Comparison between 4.0-mm stainless steel and 4.75-mm titanium alloy single-rod spinal instrumentation for anterior thoracoscopic scoliosis surgery.

    Science.gov (United States)

    Yoon, Seung Hwan; Ugrinow, Valerie L; Upasani, Vidyadhar V; Pawelek, Jeff B; Newton, Peter O

    2008-09-15

    Retrospective review of a consecutive, single surgeon case series. To compare minimum 2-year postoperative outcomes between 4.0-mm stainless steel and 4.75-mm titanium alloy single-rod anterior thoracoscopic instrumentation for the treatment of thoracic idiopathic scoliosis. Advances in anterior thoracoscopic spinal instrumentation for scoliosis have attempted to mitigate the postoperative complications of rod failure, pseudarthrosis, and deformity progression. Biomechanical data suggest that the 4.75-mm titanium construct has a lower risk of fatigue failure compared to the 4.0-mm stainless steel construct. Sixty-four consecutive anterior thoracoscopic spinal instrumentation cases in patients with thoracic scoliosis performed by a single surgeon and with minimum 2-year follow-up were retrospectively reviewed. The first 34 cases used a 4.0-mm stainless steel (SS) construct, whereas the subsequent 30 cases used a 4.75-mm titanium (Ti) alloy instrumentation system. The first 10 SS cases and the first 5 Ti cases were excluded from the statistical comparison to account for a potential learning curve effect. A multivariate analysis of variance (P 0.13). The average follow-up in the SS group was, however, significantly longer than in the Ti group (4.0 +/- 1.4 years vs. 2.3 +/- 1.0 years; P = 0.001). Preop main thoracic Cobb angles were similar between the 2 groups (P = 0.62); however, the 2-year main thoracic Cobb was significantly smaller (P = 0.03) and the 2-year percent correction was significantly greater in the Ti group (P = 0.03). Five patients (21%) in the SS group had a pseudarthrosis, 3 (13%) experienced rod failure, and 2 (8%) required a revision posterior spinal fusion. In the Ti group, 2 patients (8%) had a pseudarthrosis, and no patient experienced rod failure or required a revision procedure. Although the average follow-up in the Ti group was significantly shorter than in the SS group, the 4.75-mm titanium alloy construct resulted in improved maintenance of

  4. Comparison of galvanic corrosion potential of metal injection molded brackets to that of conventional metal brackets with nickel-titanium and copper nickel-titanium archwire combinations.

    Science.gov (United States)

    Varma, D Praveen Kumar; Chidambaram, S; Reddy, K Baburam; Vijay, M; Ravindranath, D; Prasad, M Rajendra

    2013-05-01

    The aim of the study is to investigate the galvanic corrosion potential of metal injection molding (MIM) brackets to that of conventional brackets under similar in vitro conditions with nickel-titanium and copper nickel-titanium archwires. Twenty-five maxillary premolar MIM stainless steel brackets and 25 conventional stainless steel brackets and archwires, 0.16 inch, each 10 mm length, 25 nickeltitanium wires, 25 copper nickel-titanium wires were used. They were divided into four groups which had five samples each. Combination of MIM bracket with copper nickel-titanium wire, MIM bracket with nickel-titanium wire and conventional stainless steel brackets with copper nickel-titanium wire and conventional stainless steel brackets with nickel-titanium wires which later were suspended in 350 ml of 1 M lactic acid solution media. Galvanic corrosion potential of four groups were analyzed under similar in vitro conditions. Precorrosion and postcorrosion elemental composition of MIM and conventional stainless steel bracket by scanning electron microscope (SEM) with energy dispersive spectroscope (EDS) was done. MIM bracket showed decreased corrosion susceptibility than conventional bracket with copper nickeltitanium wire. Both MIM and conventional bracket showed similar corrosion resistance potential in association with nickel-titanium archwires. It seems that both brackets are more compatible with copper nickel-titanium archwires regarding the decrease in the consequences of galvanic reaction. The EDS analysis showed that the MIM brackets with copper nickel-titanium wires released less metal ions than conventional bracket with copper nickeltitanium wires. MIM brackets showed decreased corrosion susceptibility, copper nickel-titanium archwires are compatible with both the brackets than nickel-titanium archwires. Clinically MIM and conventional brackets behaved more or less similarly in terms of corrosion resistance. In order to decrease the corrosion potential of MIM

  5. The combined action of UV irradiation and chemical treatment on the titanium surface of dental implants

    Energy Technology Data Exchange (ETDEWEB)

    Spriano, Silvia [Politecnico di Torino, Department of Applied Science and Technology, Corso Duca degli Abruzzi, 24-10129 Torino (Italy); Ferraris, Sara, E-mail: sara.ferraris@polito.it [Politecnico di Torino, Department of Applied Science and Technology, Corso Duca degli Abruzzi, 24-10129 Torino (Italy); Bollati, Daniele; Morra, Marco; Cassinelli, Clara [Nobil Bio Ricerche, Portacomaro (Italy); Lorenzon, Giorgio [Centro Chirurgico, Via Mallonetto, 47, 10032, Brandizzo Torino (Italy)

    2015-09-15

    Highlights: • A combined UV irradiation and H{sub 2}O{sub 2} treatment was applied to titanium surfaces. • A thin, homogeneous, not porous, crack-free and bioactive oxide layer was obtained. • The process significantly improves the biological response of titanium surfaces. • A clinical case demonstrates the effectiveness of the proposed treatment. - Abstract: The purpose of this paper is to describe an innovative treatment for titanium dental implants, aimed at faster and more effective osteointegration. The treatment has been performed with the use of hydrogen peroxide, whose action was enhanced by concomitant exposure to a source of ultraviolet light. The developed surface oxide layer was characterized from the physical and chemical points of view. Moreover osteoblast-like SaOS2 cells were cultured on treated and control titanium surfaces and cell behavior investigated by scanning electron microscope observation and gene expression measurements. The described process produces, in only 6 min, a thin, homogeneous, not porous, free of cracks and bioactive (in vitro apatite precipitation) oxide layer. High cell density, peculiar morphology and overexpression of several genes involved with osteogenesis have been observed on modified surfaces. The proposed process significantly improves the biological response of titanium surfaces, and is an interesting solution for the improvement of bone integration of dental implants. A clinical application of the described surfaces, with a 5 years follow-up, is reported in the paper, as an example of the effectiveness of the proposed treatment.

  6. Standard Practice for Conducting Irradiations at Accelerator-Based Neutron Sources

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    1996-01-01

    1.1 This practice covers procedures for irradiations at accelerator-based neutron sources. The discussion focuses on two types of sources, namely nearly monoenergetic 14-MeV neutrons from the deuterium-tritium T(d,n) interaction, and broad spectrum neutrons from stopping deuterium beams in thick beryllium or lithium targets. However, most of the recommendations also apply to other types of accelerator-based sources, including spallation neutron sources (1). Interest in spallation sources has increased recently due to their proposed use for transmutation of fission reactor waste (2). 1.2 Many of the experiments conducted using such neutron sources are intended to simulate irradiation in another neutron spectrum, for example, that from a DT fusion reaction. The word simulation is used here in a broad sense to imply an approximation of the relevant neutron irradiation environment. The degree of conformity can range from poor to nearly exact. In general, the intent of these simulations is to establish the fundam...

  7. Striated muscle microvascular response to silver implants: A comparative in vivo study with titanium and stainless steel.

    Science.gov (United States)

    Kraft, C N; Hansis, M; Arens, S; Menger, M D; Vollmar, B

    2000-02-01

    Local microvascular perfusion is the primary line of defense of tissue against microorganisms and plays a considerable role in reparative processes. The impairment of the microcirculation by a biomaterial may therefore have profound consequences. Silver is known to have excellent antimicrobial activity and, although regional and systemic toxic effects have been described, silver is regularly discussed as an implant material in bone surgery. Because little is known about the influence of silver implants on the adjacent host tissue microvasculature, we studied in vivo nutritive perfusion and leukocytic response, and compared these results with those of the conventionally used materials titanium and stainless steel. Using the hamster dorsal skinfold chamber preparation and intravital microscopy, the implantation of a commercially pure silver sample led to a distinct and persistent activation of leukocytes combined with a marked disruption of the microvascular endothelial integrity, massive leukocyte extravasation, and considerable venular dilation. Whereas animals with stainless-steel implants showed a moderate increase in these parameters with a tendency to recuperate, titanium implants caused only a transient increase of leukocyte-endothelial cell interaction within the first 120 min and no significant change in macromolecular leakage, leukocyte extravasation and venular diameter. After 3 days, five of six preparations with silver samples showed severe inflammation and massive edema. Thus, the use of silver as an implant material should be critically judged despite its bactericidal properties. The implant material titanium seems to be well tolerated by the local vascular system and currently represents the golden standard. Copyright 2000 John Wiley & Sons, Inc.

  8. Neutron irradiation effects in advanced superconductors

    International Nuclear Information System (INIS)

    Yoshida, H.; Kodaka, H.; Miyata, K.; Hayashi, Y.; Atobe, K.

    1988-01-01

    This paper reports the effects of neutron irradiation on superconducting transitions studied by susceptibility and resistivity measurements for A15 type compounds, Laves-phase compounds and oxide superconductors. For A15 superconductors, the transition temperature (T c ) decreased with increasing neutron fluence and showed large drop started at about 5 x 10 18 n/cm 2 (E > 0.1 MeV). Post-irradiation annealing gave recovery of T c , but the behaviors were different for the materials with different composition and microstructure. The Laves-phase compounds showed less degradation than the A15 superconductors. For oxide superconductors very sensitive transition change was observed, including the radiation-induced superconductivity

  9. Effect of microstructure on radiation induced segregation and depletion in ion irradiated SS316 steel

    International Nuclear Information System (INIS)

    Jin, Hyung Ha; Kwon, Sang Chul; Kwon, Jun Hyun

    2011-01-01

    Irradiation assisted stress corrosion cracking (IASCC), void swelling and irradiation induced hardening are caused by change of characteristics of material by neutron irradiation, stress state of material and environmental situation. It has been known that chemical compositions varies at grain boundary (GB) significantly with fluence level and the depletion of Cr element at GB has been considered as one of important factors causing material degradation, especially, IASCC in austenitic stainless steel. However, experimental results of IASCC under PWR condition were directly not connected with Cr depletion phenomenon by neutron irradiation. Because the mechanism of IASCC under PWR has not yet been clearly understood in spite of many energetic researches, fundamental researches about radiation induced segregation and depletion in irradiated austenitic stainless steels have been attracted again. In this work, an effect of residual microstructure on radiation induced segregation and depletion of alloy elements at GB was investigated in ion irradiated SS316 steel using transmission electron microscope (TEM) with energy dispersive spectrometer (EDS)

  10. Change in properties of superconducting magnet materials by fusion neutron irradiation

    International Nuclear Information System (INIS)

    Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

    2007-01-01

    A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

  11. [Susceptibility to infections and behavior of stainless steel : Comparison with titanium implants in traumatology].

    Science.gov (United States)

    Haubruck, Patrick; Schmidmaier, Gerhard

    2017-02-01

    Despite modern treatment options, implant-associated infections (IAI) remain a severe and challenging complication in the treatment of trauma patients. Almost 30 years after the introduction of implants made of titanium alloy into the treatment of trauma patients, there is still no uniform consensus regarding the clinical benefit of titanium alloy in the context of patients with IAI. We sought to determine if implants made of titanium alloy have been proven to be less susceptible regarding IAI in contrast to implants made of stainless steel. A review of the current literature on IAI in association with the utilized implant material was conducted. Relevant articles from the years 1995 to 2016 were searched in the PubMed database. A total of 183 articles were identified and all abstracts were reviewed for relevance. A total of 14 articles met the inclusion criteria and were stratified according to the level of evidence and furthermore evaluated regarding the influence of the implant material on IAI. Considerable debate remains concerning the influence of the implant material on the susceptibility to IAI; however, the available literature shows that despite slight tendencies, there is no proof of titanium alloy being favorable in the susceptibility to IAI. Furthermore, the literature shows that the design of plates for osteosynthesis might influence IAI. In particular, plates that cause less soft tissue damage and preserve perfusion of the periosteum proved to be beneficial regarding IAI.

  12. Radiation-induced grain boundary segregation in austenitic stainless steels

    International Nuclear Information System (INIS)

    Bruemmer, S.M.; Charlot, L.A.; Vetrano, J.S.; Simonen, E.P.

    1994-11-01

    Radiation-induced segregation (RIS) to grain boundaries in Fe-Ni-Cr-Si stainless alloys has been measured as a function of irradiation temperature and dose. Heavy-ion irradiation was used to produce damage levels from 1 to 20 displacements per atom (dpa) at temperatures from 175 to 550 degrees C. Measured Fe, Ni, and Cr segregation increased sharply with irradiation dose (from G to 5 dpa) and temperature (from 175 to about 350 degrees C). However, grain boundary concentrations did not change significantly as dose or temperatures were further increased. Although interfacial compositions were similar, the width of radiation-induced enrichment or depletion profiles increased consistently with increasing dose or temperature. Impurity segregation (Si and P) was also measured, but only Si enrichment appeared to be radiation-induced. Grain boundary Si peaked at levels approaching 10 at% after irradiation doses to 10 dpa at an intermediate temperature of 325 degrees C. No evidence of grain boundary silicide precipitation was detected after irradiation at any temperature. Equilibrium segregation of P was measured in the high-P alloys, but interfacial concentration did not increase with irradiation exposure. Comparisons to reported RIS in neutron-irradiated stainless steels revealed similar grain boundary compositional changes for both major alloying and impurity elements

  13. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    International Nuclear Information System (INIS)

    Jones, R.H.; Gelles, D.S.

    1993-08-01

    Ferritic-martensitic stainless steels exhibit radiation stability and stress corrosion resistance that make them attractive replacement materials for austenitic stainless steels for in-core applications. Recent radiation studies have demonstrated that 9% Cr ferritic/martensitic stainless steel had less than a 30C shift in ductile-to-brittle transition temperature (DBTT) following irradiation at 365C to a dose of 14 dpa. These steels also exhibit very low swelling rates, a result of the microstructural stability of these alloys during radiation. The 9 to 12% Cr alloys to also exhibit excellent corrosion and stress corrosion resistance in out-of-core applications. Demonstration of the applicability of ferritic/martensitic stainless steels for in-core LWR application will require verification of the irradiation assisted stress corrosion cracking behavior, measurement of DBTT following irradiation at 288C, and corrosion rates measurements for in-core water chemistry

  14. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Directory of Open Access Journals (Sweden)

    Hu J.-P.

    2016-01-01

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

  15. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

    2014-05-23

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

  16. Hydrogen in titanium alloys

    International Nuclear Information System (INIS)

    Wille, G.W.; Davis, J.W.

    1981-04-01

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

  17. In Vivo Evaluation of Immediately Loaded Stainless Steel and Titanium Orthodontic Screws in a Growing Bone

    OpenAIRE

    Gritsch, Kerstin; Laroche, Norbert; Bonnet, Jeanne-Marie; Exbrayat, Patrick; Morgon, Laurent; Rabilloud, Muriel; Grosgogeat, Brigitte

    2013-01-01

    The present work intends to evaluate the use of immediate loaded orthodontic screws in a growing model, and to study the specific bone response. Thirty-two screws (half of stainless steel and half of titanium) were inserted in the alveolar bone of 8 growing pigs. The devices were immediately loaded with a 100 g orthodontic force. Two loading periods were assessed: 4 and 12 weeks. Both systems of screws were clinically assessed. Histological observations and histomorphometric analysis evaluate...

  18. Synergistic effects of neutron and gamma ray irradiation of a commercial CHMOS microcontroller

    International Nuclear Information System (INIS)

    Xiao-Ming, Jin; Ru-Yu, Fan; Wei, Chen; Dong-Sheng, Lin; Shan-Chao, Yang; Xiao-Yan, Bai; Yan, Liu; Xiao-Qiang, Guo; Gui-Zhen, Wang

    2010-01-01

    This paper presents the experimental results of a combined irradiation environment of neutron and gamma rays on 80C196KC20, which is a 16-bit high performance member of the MCS96 microcontroller family. The electrical and functional tests were made in three irradiation environments: neutron, gamma rays, combined irradiation of neutron and gamma rays. The experimental results show that the neutron irradiation can affect the total ionizing dose behaviour. Compared with the single radiation environment, the microcontroller exhibits considerably more severe degradation in neutron and gamma ray synergistic irradiation. This phenomenon may cause a significant hardness assurance problem. (condensed matter: structure, thermal and mechanical properties)

  19. Effect of neutron irradiation on p-type silicon

    International Nuclear Information System (INIS)

    Sopko, B.

    1973-01-01

    The possibilities are discussed of silicon isotope reactions with neutrons of all energies. In the reactions, 30 Si is converted to a stable phosphorus isotope forming n-type impurities in silicon. The above reactions proceed as a result of thermal neutron irradiation. An experiment is reported involving irradiation of two p-type silicon single crystals having a specific resistance of 2000 ohm.cm and 5000 to 20 000 ohm.cm, respectively, which changed as a result of irradiation into n-type silicon with a given specific resistance. The specific resistance may be pre-calculated from the concentration of impurities and the time of irradiation. The effects of irradiation on other silicon parameters and thus on the suitability of silicon for the manufacture of semiconductor elements are discussed. (J.K.)

  20. Structural properties and neutron irradiation effects of ceramics

    International Nuclear Information System (INIS)

    Yano, Toyohiko

    1994-01-01

    In high temperature gas-cooled reactors and nuclear fusion reactors being developed at present, various ceramics are to be used in the environment of neutron irradiation for undertaking important functions. The change of the characteristics of those materials by neutron irradiation must be exactly forecast, but it has been known that the response of the materials is different respectively. The production method of ceramics and the resulted structures of ceramics which control their characteristics are explained. The features of covalent bond and ionic bond, the synthesis of powder and the phase change by heating, sintering and sintering agent, and grain boundary phase are described. The smelling of ceramics by neutron irradiation is caused by the formation of the clusters of Frenkel defects and minute spot defects. Its restoration by annealing is explained. The defects remaining in materials after irradiation are the physical defects by flipping atoms cut due to the collision with high energy particles and the chemical defects by nuclear transformation. Some physical defects can be restored, but chemical defects are never restored. The mechanical properties of ceramics and the effect of irradiation on them, and the thermal properties of ceramics and the effect of irradiation on them are reported. (K.I.)

  1. Effect of Titanium on the Microstructure and Mechanical Properties of High-Carbon Martensitic Stainless Steel 8Cr13MoV

    Directory of Open Access Journals (Sweden)

    Wen-Tao Yu

    2016-08-01

    Full Text Available The effect of titanium on the carbides and mechanical properties of martensitic stainless steel 8Cr13MoV was studied. The results showed that TiCs not only acted as nucleation sites for δ-Fe and eutectic carbides, leading to the refinement of the microstructure, but also inhibited the formation of eutectic carbides M7C3. The addition of titanium in steel also promoted the transformation of M7C3-type to M23C6-type carbides, and consequently more carbides could be dissolved into the matrix during hot processing as demonstrated by the determination of extracted carbides from the steel matrix. Meanwhile, titanium suppressed the precipitation of secondary carbides during annealing. The appropriate amount of titanium addition decreased the size and fraction of primary carbides in the as-cast ingot, and improved the mechanical properties of the annealed steel.

  2. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    International Nuclear Information System (INIS)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

  3. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

  4. Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

    International Nuclear Information System (INIS)

    Chisela, F.

    1986-01-01

    The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

  5. Annealing of dislocation loops in neutron-irradiated copper investigated by positron annihilation

    International Nuclear Information System (INIS)

    Gauster, W.B.; Mantl, S.; Schober, T.; Triftshauser, W.

    1975-01-01

    Positron annihilation angular correlation measurements were carried out on neutron-irradiated copper as a function of annealing temperature. Two types of specimens were used: single crystals irradiated with fast neutrons, and 10 B-doped polycrystalline samples irradiated with thermal neutrons. All irradiations were at approximately 320 0 K. A structure in the annealing curve, not previously observed by other techniques, indicates that between 460 and 600 0 K the dislocation loops present after irradiation dissociate and more effective positron trapping sites are formed. (auth)

  6. Treatment of Patellar Lower Pole Fracture with Modified Titanium Cable Tension Band Plus Patellar Tibial Tunnel Steel "8" Reduction Band.

    Science.gov (United States)

    Li, Jiaming; Wang, Decheng; He, Zhiliang; Shi, Hao

    2018-01-08

    To determine the efficacy of modified titanium tension band plus patellar tendon tunnel steel 8 "reduction band" versus titanium cable tension band fixation for the treatment of patellar lower pole fracture. 58 patients with lower patella fracture were enrolled in this study, including 30 patients treated with modified titanium cable tension band plus patellar tibial tunnel wire "8" tension band internal fixation (modified group), and 28 patients with titanium cable tension band fixation. All patients were followed up for 9∼15 months with an average of 11.6 months. Knee flexion was significantly improved in the modified group than in the titanium cable tension band group (111.33 ± 13 degrees versus 98.21 ± 21.70 degrees, P = 0.004). The fracture healing time showed no significant difference. At the end of the follow-up, the improvement excellent rate was 93.33% in the modified group, and 82.14% in the titanium cable tension band group. Titanium cable tension band internal fixation loosening was found in 2 cases, including 1 case of treatment by two surgeries without loose internal fixation. The modified titanium cable tension band with "8" tension band fixation showed better efficacy for lower patella fractures than titanium cable tension band fixation.

  7. A standard fission neutron irradiation facility

    International Nuclear Information System (INIS)

    Sahasrabudhe, S.G.; Chakraborty, P.P.; Iyer, M.R.; Kirthi, K.N.; Soman, S.D.

    1979-01-01

    A fission neutron irradiation facility (FISNIF) has been set up at the thermal column of the CIRUS reactor at BARC. The spectrum and the flux have been measured using threshold detectors. The paper describes the setting up of the facility, measurement and application. A concentric cylinder containing UO 2 powder sealed inside surrounds the irradiation point of a pneumatic sample transfer system located in the thermal column of the reactor. Samples are loaded in a standard aluminium capsule with cadmium lining and transported pneumatically. A sample transfer time of 1 s can be achieved in the facility. Typical applications of the facility for studying activation of iron and sodium in fission neutrons are also discussed. (Auth.)

  8. Determination of local constitutive properties of titanium alloy matrix in boron-modified titanium alloys using spherical indentation

    International Nuclear Information System (INIS)

    Sreeranganathan, A.; Gokhale, A.; Tamirisakandala, S.

    2008-01-01

    The constitutive properties of the titanium alloy matrix in boron-modified titanium alloys are different from those of the corresponding unreinforced alloy due to the microstructural changes resulting from the addition of boron. Experimental and finite-element analyses of spherical indentation with a large penetration depth to indenter radius ratio are used to compute the local constitutive properties of the matrix alloy. The results are compared with that of the corresponding alloy without boron, processed in the same manner

  9. Impact of neutron irradiation on mechanical performance of FeCrAl alloy laser-beam weldments

    Science.gov (United States)

    Gussev, M. N.; Cakmak, E.; Field, K. G.

    2018-06-01

    Oxidation-resistant iron-chromium-aluminum (FeCrAl) alloys demonstrate better performance in Loss-of-Coolant Accidents, compared with austenitic- and zirconium-based alloys. However, further deployment of FeCrAl-based materials requires detailed characterization of their performance under irradiation; moreover, since welding is one of the key operations in fabrication of light water reactor fuel cladding, FeCrAl alloy weldment performance and properties also should be determined prior to and after irradiation. Here, advanced C35M alloy (Fe-13%Cr-5%Al) and variants with aluminum (+2%) or titanium carbide (+1%) additions were characterized after neutron irradiation in Oak Ridge National Laboratory's High Flux Isotope Reactor at 1.8-1.9 dpa in a temperature range of 195-559 °C. Specimen sets included as-received (AR) materials and specimens after controlled laser-beam welding. Tensile tests with digital image correlation (DIC), scanning electron microscopy-electron back scatter diffraction analysis, fractography, and x-ray tomography analysis were performed. DIC allowed for investigating local yield stress in the weldments, deformation hardening behavior, and plastic anisotropy. Both AR and welded material revealed a high degree of radiation-induced hardening for low-temperature irradiation; however, irradiation at high-temperatures (i.e., 559 °C) had little overall effect on the mechanical performance.

  10. Neutron irradiation of rat embryos in utero

    International Nuclear Information System (INIS)

    Vogel, H.H. Jr.

    1978-01-01

    In the rat radiation is most effective in producing congenital anomalies during the organ-forming period (days 9 to 13), which is approximately equivalent to the 14th to 50th days of human pregnancy. We have exposed female Sprague--Dawley rats on the 18th day of pregnancy to single whole-body doses of fission neutrons (20 to 150 rads). After 20 rads there was a small decrease in body weight which lasted from birth to weaning. During this period 9% of the irradiated rats died compared with 4% of the controls. After 50 rads, 65/275 (23.6%) of the rats died between birth and weaning, and the body-weight loss of the survivors was increased. After 100 rads, 62/133 (47%) died at birth or day 1 and 103/133 (77.4%) died before weaning. A large and significant decrease in body weight persisted in the survivors. After 150 rads of fission neutrons, all 95 rats died within 48 hr of birth. From cross-fostering experiments, we believe this is a direct effect of radiation on the embryos and not an indirect action through the mother or her milk. The LD 50 for the period from birth to weaning is approximately 75 rads of fission neutrons. Studies of organ weight were conducted daily for the first week after birth in an attempt to find the cause of radiation mortality. Body weight of the irradiated animals averaged only about one-half that of the controls. The liver, kidney, brain, and testes of the neutron-irradiated rats weighed significantly less than those of the controls. The weights of the spleen, lungs, duodenum, and stomach were decreased but not significantly. The bone marrow appeared depleted in the irradiated long bones, but the spleen maintained active hematopoiesis 1 to 2 months after neutron exposure

  11. Canal preparation with nickel-titanium or stainless steel instruments without the risk of instrument fracture: preliminary observations

    OpenAIRE

    Ghassan Yared

    2015-01-01

    This report introduces a novel technique that allows a safe and predictable canal negotiation, creation of a glide path and canal preparation with reciprocating nickel-titanium or stainless steel engine-driven instruments in canals where the use of rotary and the newly developed reciprocating instruments is contraindicated. In this novel technique, the instruments are used in reciprocating motion with very small angles. Hand files are not used regardless of the complexity of the canal anatomy...

  12. Thermal analysis of the APT materials irradiation samples

    International Nuclear Information System (INIS)

    Maloy, S.A.; Willcutt, G.J.; James, M.R.; Teague, J.; Diebe, D.A.; Sommer, W.F.; Ferguson, P.D.

    1998-01-01

    The accelerator production of tritium (APT) project proposes to use a 1.7 GeV, 100 mA proton beam to produce neutrons from an Inconel 718 clad tungsten target. The neutrons are multiplied and moderated in a lead/water blanket before being captured in He 3 to form tritium. In this process, the materials in the target and blanket region are exposed to a wide range of different fluxes comprised of protons and neutrons with energies into the GeV range. To investigate the effect of irradiation on the mechanical properties of candidate APT materials (Inconel 718, 316L stainless steel, Al 6061-T6, Mod 9Cr-1Mo, 304L stainless steel and Al5052-0), the APT Engineering Design and Development group fielded an extensive materials irradiation using the LANSCE (Los Alamos Neutron Science Center) accelerator, which operates at an energy of 800 MeV and a current of 1 mA. The test set-up was designed to place mechanical test specimens in locations in and near the proton beam where the environment of proton and neutron fluxes and temperatures are prototypic to those expected in the APT target/blanket (50--170 C). After irradiating for about 3,600 hours, the maximum achieved proton fluence was 4--5 x 10 21 p/cm 2 for the materials in the center of the beam. To obtain relevant data on the change in the mechanical properties with fluence, it is essential to know the temperature at which the materials were irradiated. This paper explains the method of determining the specimen temperature and reports some specific examples

  13. Experimental measurement of nuclear heating in a graphite-cantered assembly in deuterium-tritium neutron environment for the validation of data and calculation

    International Nuclear Information System (INIS)

    Kumar, A.; Youssef, M.; Abdou, M.A.

    1998-01-01

    Within the framework of the ITER Task T-218 entitled 'Shielding Blanket Neutronics Experiments', nuclear heating measurements were conducted jointly by the USA and Japan using a micro calorimetric technique in a graphite-cantered assembly. An accelerator-based D-T neutron source at JAERI was used to provide a mixed neutron and photon field. The first measurements related to direct micro calorimetric measurements in individual graphite probes along the axis. In the second set, the first graphite probe was replaced, one by one, by single probes of beryllium, aluminum, silicon, silicon carbide, titanium, vanadium, chromium, iron, stainless steel 316, nickel, copper, zirconium, niobium, molybdenum, tungsten. Analysis of the measurements has been carried out using Monte Carlo code MCNP with FENDL-1, ENDF/B-VI and MCPLIB nuclear data libraries. A comparison of calculations (C) and experiments (E) shows a C/E ratio lying in a C/E band extending from 0.9 to 1.2 for beryllium, graphite, copper, chromium, iron, nickel, 316 stainless steel, titanium, vanadium, molybdenum, niobium and tungsten. However, larger deviations from unity are seen for C/E values for silicon, zirconium, and aluminum. Though FENDL-1 and ENDF/B-VI libraries provide very close nuclear heating rates for most of the probe materials, significant divergences are seen for silicon, silicon carbide, aluminum, titanium, zirconium, niobium, and molybdenum. The divergences are traceable to differences in neutron kerma factors as well as gamma production cross-sections of these materials. (orig.)

  14. Femtosecond laser-induced periodic surface structures on steel and titanium alloy for tribological applications

    Science.gov (United States)

    Bonse, J.; Koter, R.; Hartelt, M.; Spaltmann, D.; Pentzien, S.; Höhm, S.; Rosenfeld, A.; Krüger, J.

    2014-10-01

    Laser-induced periodic surface structures (LIPSS, ripples) were generated on stainless steel (100Cr6) and titanium alloy (Ti6Al4V) surfaces upon irradiation with multiple femtosecond laser pulses (pulse duration 30 fs, central wavelength 790 nm). The experimental conditions (laser fluence, spatial spot overlap) were optimized in a sample-scanning geometry for the processing of large surface areas (5 × 5 mm2) covered homogeneously by the nanostructures. The irradiated surface regions were subjected to white light interference microscopy and scanning electron microscopy revealing spatial periods around 600 nm. The tribological performance of the nanostructured surface was characterized by reciprocal sliding against a ball of hardened steel in paraffin oil and in commercial engine oil as lubricants, followed by subsequent inspection of the wear tracks. For specific conditions, on the titanium alloy a significant reduction of the friction coefficient by a factor of more than two was observed on the laser-irradiated (LIPSS-covered) surface when compared to the non-irradiated one, indicating the potential benefit of laser surface structuring for tribological applications.

  15. Quantitative analyses of impurity silicon-carbide (SiC) and high-purity-titanium by neutron activation analyses based on k0-standardization method. Development of irradiation silicon technology in productivity using research reactor (Joint research)

    International Nuclear Information System (INIS)

    Motohashi, Jun; Takahashi, Hiroyuki; Magome, Hirokatsu; Sasajima, Fumio; Tokunaga, Okihiro; Kawasaki, Kozo; Onizawa, Koji; Isshiki, Masahiko

    2009-07-01

    JRR-3 and JRR-4 have been providing neutron-transmutation-doped silicon (NTD-Si) by using the silicon NTD process, which is a method to produce a high quality semiconductor. The domestic supply of NTD-Si is insufficient for the demand, and the market of NTD-Si is significantly growing at present. It is very important to increase achieve the production. To fulfill the requirement, we have been investigating a neutron filter, which is made of high-purity-titanium, for uniform doping. Silicon-carbide (SiC) semiconductor doped with NTD technology is considered suitable for high power devices with superior performances to conventional Si-based devices. We are very interested in the SiC as well. This report presents the results obtained after the impurity contents in the high-purity-titanium and SiC were analyzed by neutron activation analyses (NAA) using k 0 -standardization method. There were 6 and 9 impurity elements detected from the high-purity-titanium and SiC, respectively. Among those Sc from the high-purity-titanium and Fe from SiC were comparatively long half life nuclides. From the viewpoint of exposure in handling them, we need to examine the impurity control of materials. (author)

  16. Neutron self-shielding with k0-NAA irradiations

    International Nuclear Information System (INIS)

    Chilian, C.; Chambon, R.; Kennedy, G.

    2010-01-01

    A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

  17. Bacterial adhesion studies on titanium, titanium nitride and modified hydroxyapatite thin films

    Energy Technology Data Exchange (ETDEWEB)

    Jeyachandran, Y L [Department of Physics, Bharathiar University, Coimbatore 641 046, Tamil Nadu (India); Venkatachalam, S [Department of Physics, Bharathiar University, Coimbatore 641 046, Tamil Nadu (India); Karunagaran, B [Department of Physics, Bharathiar University, Coimbatore 641 046, Tamil Nadu (India); Narayandass, Sa K [Department of Physics, Bharathiar University, Coimbatore 641 046, Tamil Nadu (India); Mangalaraj, D [Department of Physics, Bharathiar University, Coimbatore 641 046, Tamil Nadu (India); Bao, C Y [West China College of Stomatology, Sichuan University, Chengdu 610041 (China); Zhang, C L [West China College of Stomatology, Sichuan University, Chengdu 610041 (China)

    2007-01-15

    A qualitative study on adhesion of the oral bacteria Porphyromonas gingivalis on titanium (Ti), titanium nitride (TiN), fluorine modified hydroxyapatite (FHA) and zinc modified FHA (Zn-FHA) thin films is investigated. Ti and TiN thin films were deposited by DC magnetron sputtering and hydroxyapatite-based films were prepared by solgel method. The crystalline structure, optical characteristics, chemical composition and surface topography of the films were studied by XRD, optical transmission, XPS, EDAX and AFM measurements. The predominant crystallite orientation in the Ti and TiN films was along (002) and (111) of hcp and cubic structures, respectively. The Ti : O : N composition ratio in the surface of the Ti and TiN films was found to be 7 : 21 : 1 and 3 : 8 : 2, respectively. The atomic concentration ratio (Zn + Ca) / P in Zn-FHA film was found to be 1.74 whereby the Zn replaced 3.2% of Ca. The rough surface feature in modified HA films was clearly observed in the SEM images and the surface roughness (rms) of Ti and TiN films was 2.49 and 3.5 nm, respectively, as observed using AFM. The film samples were sterilized, treated in the bacteria culture medium, processed and analyzed using SEM. Surface roughness of the films was found to have least influence on the bacterial adhesion. More bacteria were observed on the TiN film with oxide nitride surface layer and less number of adhered bacteria was noticed on the Ti film with native surface oxide layer and on Zn-FHA film.

  18. Bacterial adhesion studies on titanium, titanium nitride and modified hydroxyapatite thin films

    International Nuclear Information System (INIS)

    Jeyachandran, Y.L.; Venkatachalam, S.; Karunagaran, B.; Narayandass, Sa.K.; Mangalaraj, D.; Bao, C.Y.; Zhang, C.L.

    2007-01-01

    A qualitative study on adhesion of the oral bacteria Porphyromonas gingivalis on titanium (Ti), titanium nitride (TiN), fluorine modified hydroxyapatite (FHA) and zinc modified FHA (Zn-FHA) thin films is investigated. Ti and TiN thin films were deposited by DC magnetron sputtering and hydroxyapatite-based films were prepared by solgel method. The crystalline structure, optical characteristics, chemical composition and surface topography of the films were studied by XRD, optical transmission, XPS, EDAX and AFM measurements. The predominant crystallite orientation in the Ti and TiN films was along (002) and (111) of hcp and cubic structures, respectively. The Ti : O : N composition ratio in the surface of the Ti and TiN films was found to be 7 : 21 : 1 and 3 : 8 : 2, respectively. The atomic concentration ratio (Zn + Ca) / P in Zn-FHA film was found to be 1.74 whereby the Zn replaced 3.2% of Ca. The rough surface feature in modified HA films was clearly observed in the SEM images and the surface roughness (rms) of Ti and TiN films was 2.49 and 3.5 nm, respectively, as observed using AFM. The film samples were sterilized, treated in the bacteria culture medium, processed and analyzed using SEM. Surface roughness of the films was found to have least influence on the bacterial adhesion. More bacteria were observed on the TiN film with oxide nitride surface layer and less number of adhered bacteria was noticed on the Ti film with native surface oxide layer and on Zn-FHA film

  19. Fracture toughness and strength change of neutron-irradiated ceramic materials

    International Nuclear Information System (INIS)

    Dienst, W.; Zimmermann, H.

    1994-01-01

    In order to analyse the results of bending strength measurements on neutron-irradiated samples of Al 2 O 3 , AlN and SiC, fracture toughness measurements were additionally conducted. The neutron fluences concerned were mostly in the range of 0.6 to 3.2x10 26 n/m 2 at irradiation temperatures of 400 to 550 C. A fracture toughness decrease was generally observed for polycrystalline materials which, however, was considerably smaller than the reduction of the fracture strength. Exceptional increase of the fracture toughness seems typical for the effect of rather coarse irradiation defects. The irradiation-induced change of the fracture toughness of single crystal Al 2 O 3 appeared dependent on the crystallographic orientation; both reduced and increased fracture toughness after irradiation was observed. Recent results of neutron irradiation to about 2x10 25 n/m 2 at 100 C showed, that the strength decrease of various Al 2 O 3 grades sets in at (3-5)x10 24 n/m 2 and seems to be little dependent on the irradiation temperature. ((orig.))

  20. Irradiation test of borosilicate glass burnable poison

    International Nuclear Information System (INIS)

    Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

    1991-08-01

    The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

  1. Study of stress relief cracking in titanium stabilized austenitic stainless steel; Etude de la fissuration differee par relaxation d'un acier inoxydable austenitique stabilise au titane

    Energy Technology Data Exchange (ETDEWEB)

    Chabaud-Reytier, M

    1999-07-01

    The heat affected zone (HAZ) of titanium stabilised austenitic stainless steel welds (AISI 321) may exhibit a serious form of intercrystalline cracking during service at high temperature. This type of cracking, called 'stress relief cracking', is known to be due to work hardening but also to ageing: a fine and abundant intragranular Ti(C,N) precipitation appears near the fusion line and modifies the mechanical behaviour of the HAZ. This study aims to better know the accused mechanism and to succeed in estimating the risk of such cracking in welded junctions of 321 stainless steel. To analyse this embrittlement mechanism, and to assess the lifetime of real components, different HAZ are simulated by heat treatments applied to the base material which is submitted to various cold rolling and ageing conditions in order to reproduce the HAZ microstructure. Then, we study the effects of work hardening and ageing on the titanium carbide precipitation, on the mechanical (tensile and creep) behaviour of the resulting material and on its stress relief cracking sensitivity. It is shown that work hardening is the main parameter of the mechanism and that ageing do not favour crack initiation although it leads to titanium carbide precipitation. The role of this precipitation is also discussed. Moreover, a creep damage model is identified by a local approach to fracture. Materials sensitive to stress relief cracking are selected. Then, creep tests are carried out on notched bars in order to quantify the intergranular damage of these different materials; afterwards, these measurements are combined with calculated mechanical fields. Finally, it is shown that the model gives good results to assess crack initiation for a compact tension (CT) specimen during relaxation tests, as well as for a notched tubular specimen tested at 600 deg. C under a steady torque. (author)

  2. Shielding of a neutron irradiator with {sup 241}Am-Be source

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

  3. Investigation of neutron fluence using fluence monitors for irradiation test at WWR-K

    International Nuclear Information System (INIS)

    Romanova, N.K.; Takemoto, N.

    2013-01-01

    Irradiation test of a Si ingot is planned using WWR-K in Institute of Nuclear Physics Republic of Kazakhstan (INP RK) to develop an irradiation technology for Si semiconductor production by Neutron Transmutation Doping (NTD) method in the framework of an international cooperation between INP RK and Japan Atomic Energy Agency (JAEA), Japan. It is possible to irradiate the Si ingot of 6 inch in diameter at the K-23 irradiation channel in the WWR-K. The preliminary irradiation test using 4 Al ingots was performed to evaluate the actual neutronic irradiation field at the K-23 channel in the WWR-K. Each Al ingot has the same dimension as the Si ingot, and 15 fluence monitors are equipped in it. Iron wire and aluminum-cobalt wire are inserted into them, and it is possible to evaluate both fast and thermal neutron fluxes by measurement of these radiation activities after irradiation. This report described the results of the preliminary irradiation test and the neutronic calculations by Monte Carlo method in order to evaluate the neutronic irradiation field in the irradiation position for the silicon ingot at the channel in the WWR-K. (authors)

  4. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  5. Neutron irradiation of sapphire for compressive strengthening. II. Physical properties changes

    Energy Technology Data Exchange (ETDEWEB)

    Regan, Thomas M. E-mail: thomas_regan@uml.edu; Harris, Daniel C. E-mail: harrisdc@navair.navy.mil; Blodgett, David W.; Baldwin, Kevin C.; Miragliotta, Joseph A.; Thomas, Michael E.; Linevsky, Milton J.; Giles, John W.; Kennedy, Thomas A.; Fatemi, Mohammad; Black, David R.; Lagerloef, K. Peter D

    2002-01-01

    Irradiation of sapphire with fast neutrons (0.8-10 MeV) at a fluence of 10{sup 22}/m{sup 2} increased the c-axis compressive strength and the c-plane biaxial flexure strength at 600 deg. C by a factor of {approx}2.5. Both effects are attributed to inhibition of r-plane twin propagation by damage clusters resulting from neutron impact. The a-plane biaxial flexure strength and four-point flexure strength in the c- and m-directions decreased by 10-23% at 600 deg. C after neutron irradiation. Neutron irradiation had little or no effect on thermal conductivity, infrared absorption, elastic constants, hardness, and fracture toughness. A featureless electron paramagnetic resonance signal at g=2.02 was correlated with the strength increase: This signal grew in amplitude with increasing neutron irradiation, which also increased the compressive strength. Annealing conditions that reversed the strengthening also annihilated the g=2.02 signal. A signal associated with a paramagnetic center containing two Al nuclei was not correlated with strength. Ultraviolet and visible color centers also were not correlated with strength in that they could be removed by annealing at temperatures that were too low to reverse the compressive strengthening effect of neutron irradiation.

  6. Total cross-sections assessment of neutron reaction with stainless steel SUS-310 contained in various nuclear data files

    International Nuclear Information System (INIS)

    Suwoto

    2002-01-01

    The integral testing of neutron cross-sections for Stainless Steel SUS-310 contained in various nuclear data files have been performed. The shielding benchmark calculations for Stainless Steel SUS-310 has been analysed through ORNL-Broomstick Experiment calculation which performed by MAERKER, R.E. at ORNL - USA ( 1) . Assessment with JENDL-3.1, JENDL-3.2, ENDF/B-IV, ENDF/B-VI nuclear data files and data from GEEL have also been carried out. The overall calculation results SUS-310 show in a good agreement with the experimental data, although, underestimate results appear below 3 MeV for all nuclear data files. These underestimation tendencies clearly caused by presented of iron nuclide which more than half in Stainless Steel compound. The total neutron cross-sections of iron nuclide contained in various nuclear data files relatively lower on that energy ranges

  7. Neutron and gamma irradiation effects on power semiconductor switches

    Science.gov (United States)

    Schwarze, G. E.; Frasca, A. J.

    1990-01-01

    The performance characteristics of high power semiconductor switches subjected to high levels of neutron fluence and gamma dose must be known by the designer of the power conditioning, control and transmission subsystem of space nuclear power systems. Location and the allowable shielding mass budget will determine the level of radiation tolerance required by the switches to meet performance and reliability requirements. Neutron and gamma ray interactions with semiconductor materials and how these interactions affect the electrical and switching characteristics of solid state power switches is discussed. The experimental measurement system and radiation facilities are described. Experimental data showing the effects of neutron and gamma irradiation on the performance characteristics are given for power-type NPN Bipolar Junction Transistors (BJTs), and Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs). BJTs show a rapid decrease in gain, blocking voltage, and storage time for neutron irradiation, and MOSFETs show a rapid decrease in the gate threshold voltage for gamma irradiation.

  8. 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.

  9. An investigation of neutron irradiation test on superplastic zirconia-ceramic materials

    International Nuclear Information System (INIS)

    Shibata, Taiju; Ishihara, Masahiro; Baba, Shinichi; Hayashi, Kimio

    2000-05-01

    A neutron irradiation test on superplastic ceramic materials at high temperature has been proposed as an innovative basic research on high-temperature engineering using the High Temperature Engineering Test Reactor (HTTR). For the effective execution of the test, we reviewed the superplastic deformation mechanism of ceramic materials and discussed neutron irradiation effects on the superplastic deformation process of stabilized Tetragonal Zirconia Polycrystal (TZP), which is a representative superplastic ceramic material. As a result, we pointed out that the decrease in the activation energy for superplastic deformation is expected by the radiation-enhanced diffusion. We selected a fast neutron fluence of 5x10 20 n/cm 2 and an irradiation temperature of about 600degC as test conditions for the first irradiation test on TZP and decided to perform a preliminary irradiation test by the Japan Materials Testing Reactor (JMTR). Moreover, we estimated the radioactivity of irradiated TZP and indicated that it is in the order of 10 10 Bq/g (about 0.3 Ci/g) immediately after irradiation to a thermal neutron fluence of 3x10 20 n/cm 2 and that it decays to about 1/100 in a year. (author)

  10. Fusion neutron irradiation of Ni-Si alloys at high temperature*1

    Science.gov (United States)

    Huang, J. S.; Guinan, M. W.; Hahn, P. A.

    1988-07-01

    Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

  11. Aspects of Low Temperature Irradiation in Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Brune, D.

    1968-08-01

    Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined

  12. Aspects of Low Temperature Irradiation in Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Brune, D

    1968-08-15

    Neutron irradiation of the sample while frozen in a cooling device inserted in a reactor channel has been carried out in the analysis of iodine in aqueous samples as well as of mercury in biological tissue and water. For the simultaneous irradiation of a large number of aqueous solutions the samples were arranged in a suitable geometry in order to avoid mutual flux perturbation effects. The influence of the neutron temperature on the activation process has been discussed. Potential applications of the low temperature irradiation technique are outlined.

  13. Verification of neutron irradiation on S/G tube materials

    International Nuclear Information System (INIS)

    Kang, Byoung Hwi; Lee, S. K.; Jang, D. Y.; Jo, K. H.

    2010-12-01

    The fluence monitors were fabricated with metal wires of the purity ≥ 99.9%, whose dimensions were 0.1mm diameter, about 3mm length, and around 150-200 μg mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position. The measured neutron fluences were compared to the calculated ones. (Errors ≤ 35%)

  14. Corrosion and microstructural aspects of dissimilar joints of titanium and type 304L stainless steel

    International Nuclear Information System (INIS)

    Mudali, U. Kamachi.; Ananda Rao, B.M.; Shanmugam, K.; Natarajan, R.; Raj, Baldev

    2003-01-01

    To link titanium and zirconium metal based (Ti, Zr-2, Ti-5%Ta, Ti-5%Ta-1.8Nb) dissolver vessels containing highly radioactive and concentrated corrosive nitric acid solution to other nuclear fuel reprocessing plant components made of AISI type 304L stainless steel (SS), high integrity and corrosion resistant dissimilar joints between them are necessary. Fusion welding processes produce secondary precipitates which dissolve in nitric acid, and hence solid-state processes are proposed. In this work, various dissimilar joining processes available for producing titanium-304L SS joints with adequate strength, ductility and corrosion resistance for this critical application are highlighted. Developmental efforts made at IGCAR, Kalpakkam are outlined. The possible methods and the microstructural-metallurgical properties of the joints along with corrosion results obtained with three phase (liquid, vapour, condensate) corrosion testing are discussed. Based on the results, dissimilar joint produced by the explosive joining process was adopted for plant application

  15. 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.

  16. Effect of stress during neutron irradiation on the microstructure of type 316 stainless steel

    International Nuclear Information System (INIS)

    Brager, H.R.; Garner, F.A.; Guthrie, G.L.

    1976-04-01

    A transmission electron microscopy (TEM) examination was performed on solution annealed and 20 percent cold-worked Type 316 stainless steel specimens stressed during irradiation at 500 0 C in EBR-II. Hoop stress levels ranged from 0 to 327 MN/m 2 (47,300 psi) and fluences between 2.0 and 3.0 x 10 22 n/cm 2 (E greater than 0.1 MeV). Data confirm that applied tensile stresses enhance swelling in the solution annealed steel. The number densities of both voids and Frank loops were sensitive to the stress environment. Total swelling in the annealed material increased with stress, but not in direct proportion to the increased void nucleation. While the effect of cold working was to suppress swelling, the nucleation and growth of Frank loops was unaffected by the cold worked microstructure. The individual planar loop densities within any one specimen were quite sensitive to the magnitude of the stress component normal to the loop plane, while the total loop number density was sensitive to a smaller degree of the magnitude of the hydrostatic stress level. The number and size distribution of the loop populations were unaffected by the planar shear stress components, but the mean loop sizes were found to be limited by the probability of loop intersection with dislocations and loops. The stress dependence of void and loop densities allowed determination of the critical nuclei sizes, approximately sixteen vacancies for voids and six atoms for loops. Many observations were made on the probable creep mechanisms. Both dislocation and void microstructures evolved in a consistent stress-dependent manner, giving support to models which predict a coupling of the swelling and irradiation creep phenomena through the stress environment. 13 figures, 3 tables

  17. Bloodcompatibility improvement of titanium oxide film modified by phosphorus ion implantation

    International Nuclear Information System (INIS)

    Yang, P.; Leng, Y.X.; Zhao, A.S.; Zhou, H.F.; Xu, L.X.; Hong, S.; Huang, N.

    2006-01-01

    Our recent investigation suggested that Ti-O thin film could be a newly developed antithrombotic material and its thromboresistance could be related to its physical properties of wide gap semiconductor. In this work, titanium oxide film was modified by phosphorus ion implantation and succeeding vacuum annealing. RBS were used to investigate phosphorus distribution profile. Contact angle test results show that phosphorus-doped titanium oxide film becomes more hydrophilic after higher temperature annealing, while its electric conductivity increases. Antithrombotic property of phosphorus-doped titanium oxide thin films was examined by clotting time and platelet adhesion tests. The results suggest that phosphorus doping is an effective way to improve the bloodcompatibility of titanium oxide film, and it is related to the changes of electron structure and surface properties caused by phosphorus doping

  18. Void swelling of proton-irradiated stainless steel at large displacement levels

    International Nuclear Information System (INIS)

    Kumar, A.; Garner, F.A.

    1982-01-01

    The purpose of this study is to determine whether saturation of void swelling in AISI 316 stainless steel can be made to occur at any level relevant to engineering design and to decide whether saturation is sensitive to irradiation variables such as helium/dpa ratio or simulation artifacts such as injected interstitials

  19. 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

  20. Neutron irradiation induced amorphization of silicon carbide

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Development of irradiation rig in HTTR and dosimetry method. I-I type irradiation equipment

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Taiju; Kikuchi, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi; Ogura, Kazutomo [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated, helium gas-cooled test reactor with a maximum power of 30 MW. The HTTR aims not only to establish and upgrade the technological basis for the HTGRs but also to perform the innovative basic research on high temperature engineering with high temperature irradiation fields. It is planned that the HTTR is used to perform various engineering tests such as the safety demonstration test, high temperature test operation and irradiation test with large irradiation fields at high temperatures. This paper describes the design of the I-I type irradiation equipment developed as the first irradiation rig for the HTTR and does the planned dosimetry method at the first irradiation test. It was developed to perform in-pile creep test on a stainless steel with large standard size specimens in the HTTR. It can give great loads on the specimens stably and can control the irradiation temperature precisely. The in-core creep properties on the specimens are measured by newly developed differential transformers and the irradiation condition in the core is monitored by thermocouples and self-powered neutron detectors (SPNDs), continuously. The irradiated neutron fluence is assessed by neutron fluence monitors of small metallic wires after the irradiation. The obtained data at the first irradiation test can strongly be contributed to upgrade the technological basis for the HTGRs, since it is the first direct measurement of the in-core irradiation environments of the HTTR. (author)

  2. Characterization of defect accumulation in neutron-irradiated Mo by positron annihilation spectroscopy

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Li, Meimei; Snead, L.L.

    2008-01-01

    Positron annihilation lifetime spectroscopy measurements were performed on neutron-irradiated low carbon arc cast Mo. Irradiation took place in the high flux isotope reactor, Oak Ridge National Laboratory, at a temperature of 80 +/- 10 degrees C. Neutron fluences ranged from 2 x 10(21) to 8 x 10(......, as predicted by molecular dynamics simulations. (C) 2008 Elsevier B.V. All rights reserved....... at a very low-dose of similar to 10(-4) dpa. The average size of the cavities did not change significantly with dose, in contrast to neutron-irradiated bcc Fe where cavity sizes increased with increasing dose. It is suggested that the in-cascade vacancy clustering may be significant in neutron-irradiated Mo...

  3. Colony form variation of Bacillus pumilus E601 after cultured and neutron irradiation

    International Nuclear Information System (INIS)

    Chen Xiaoming; Wei Baoli; Zhang Jianguo

    2008-01-01

    The distribution of two colony forms of Bacillus pumilus E601 and the effect of neutron irradiation on the colony form were reported. The translucent and opaque colonies were cultured several generations to observe the proportion of two form colonies. The spores of opaque colony were irradiated at 80, 800 and 2000 Gy of fast neutron from CFBR-II pulse pile, and the survivors of opaque colony were irradiated again at the same doses. The results showed that: (1) Bacillus pumilus E601 observed two types of colony form: translucent and opaque colony; (2) the translucent colony could produce both translucent and opaque colonies in equal, while the opaque colony couldn't produce translucent colony generally; (3) neutron irradiation could affect the colony form distribution. The ratio of survival translucent colony was increased with the increase of the first neutron irradiation doses, and the second neutron irradiation also increased the ratio of translucent colony. It was concluded that the instability of translucent colony was the main reason to produce two colony forms of Bacillus pumilus E601. The strain of translucent colony had a stronger ability to resist neutron irradiation than the opaque colony. (authors)

  4. Further study of the glassy low-temperature properties of irradiated crystalline quartz: neutron and electron irradiation

    International Nuclear Information System (INIS)

    Laermans, C.; Daudin, B.

    1979-01-01

    Recently it has been shown that a quartz crystal after light fast neutron irradiation shows low temperature hypersonic properties which are similar to those found in glasses although the sample was still crystalline. Additional measurements have been carried out in the neutron-irradiated sample and a sample irradiated with high energy electrons has also been investigated. (Fast neutron dose 6 x 10 18 n/cm 2 , 2 MeV electron dose 3 x 10 19 e/cm 2 ). A magnetic field up to 1.5 T was found to have no influence in the hypersonic saturation behaviour of the neutron-irradiated sample (9 GHz, 1.65 K) and thermal conductivity measurements are consistent with a number of two level systems (2 LS) an order of magnitude lower than in vitreous silica as found before. Low temperature hypersonic measurements as a function of acoustic intensity and temperature as well as thermal conductivity measurements give no evidence for the presence of 2 LS in the electron irradiated sample. Considering the damage created in both samples this indicates that 2 LS are probably not related to point defects

  5. Studsvik thermal neutron facility

    International Nuclear Information System (INIS)

    Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

    1992-01-01

    The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

  6. Simulation analysis of radiation fields inside phantoms for neutron irradiation

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.

    2007-01-01

    Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)

  7. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

  8. Tensile and fracture toughness test results of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1998-01-01

    Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

  9. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  10. Irradiation creep of solution annealed and coldworked 316 stainless steel

    International Nuclear Information System (INIS)

    Boutard, J.L.; Carteret, Y.; Cauvin, R.; Guerin, Y.; Maillard, A.

    1983-04-01

    Because SA and CW 316 stainless steels were used as standard cladding material, a lot of plastic strain data is now avalaible. Most of it is published and analyzed in term of an irradiation creep modulus A defined as the ratio of the equivalent plastic strain to the product of the equivalent stress by the dose. In fact the experimental data and the theoretical analysis of the in-pile deformation mechanisms show a more complicated situation. The purpose of this paper is to reanalyze our results taking into account this situation. This approach is divided in two parts: 1) the high temperature range (T>=450 0 C) where data come from irradiated pins; 2) the low temperature range (T 0 C) where results come from pressurized tubes irradiated in experimental rigs

  11. On the recovery of neutron irradiation defects of some metals and alloys

    International Nuclear Information System (INIS)

    Mohamed, H.G.; Matta, M.K.

    2001-01-01

    This work deals with the recovery of mechanical properties of neutron irradiated material to the pre-irradiating values. Rate of migration of defects responsible for radiation hardening and those inducing radiation embrittlement is analyzed. Role of crystalline structure is also studied. Materials of FCC crystal structure used in these investigations are pure Cu, Cu-5 at. % , Al, Cu-5 at. % Si, some Ni base binary alloys and some austenitic stainless steels mainly of AISI types 304 and 316. Among materials of BCC crystalline structure Fe-6 wt % Cr alloy is used. Alloys with CPH structure used in the present investigations are Zr-l wt. % Nb and Mg - 4.8 wt % Li alloys. History of material is studied such as cold worked state and annealed condition. Character of alloying elements and their amounts were of interest in this study. The result showed that the higher the percentage radiation hardening, the slower is the migration of radiation defects. Irradiated pure metals recovered at a higher temperature than alloys. Cold work accelerated the migration of radiation defects. The amount of alloying elements had little effect on the recovery temperatures. Character of solute alloying elements (substitutional or interstitial) revealed sensitive effect on the migration of radiation defects. Rate of migration of defects causing hardening can be different from those causing embrittlement. (author)

  12. Repair-welding technology of irradiated materials - WIM project

    International Nuclear Information System (INIS)

    Nakata, K.; Oishi, M.

    1998-01-01

    A new project on the development of repair-welding technology for core internals and reactor (pressure) vessel, consigned by the Ministry of International Trade and Industry (MITI), has been started from October 1997. The objective of the project is classified into three points as follows: (1) to develop repair-welding techniques for neutron irradiated materials, (2) to prove the availability of the techniques for core internals and reactor (pressure) vessel, and (3) to recommend the updated repair-welding for the Technical Rules and Standards. Total planning, neutron irradiation, preparation of welding equipment are now in progress. The materials are austenitic stainless steels and a low alloy steel. Neutron irradiation is performed using test reactors. In order to suppress the helium aggregation along grain boundaries, low heat input welding techniques, such as laser, low heat input TIG and friction weldings, will be applied. (author)

  13. Using sewage sludge pyrolytic gas to modify titanium alloy to obtain high-performance anodes in bio-electrochemical systems

    Science.gov (United States)

    Gu, Yuan; Ying, Kang; Shen, Dongsheng; Huang, Lijie; Ying, Xianbin; Huang, Haoqian; Cheng, Kun; Chen, Jiazheng; Zhou, Yuyang; Chen, Ting; Feng, Huajun

    2017-12-01

    Titanium is under consideration as a potential stable bio-anode because of its high conductivity, suitable mechanical properties, and electrochemical inertness in the operating potential window of bio-electrochemical systems; however, its application is limited by its poor electron-transfer capacity with electroactive bacteria and weak ability to form biofilms on its hydrophobic surface. This study reports an effective and low-cost way to convert a hydrophobic titanium alloy surface into a hydrophilic surface that can be used as a bio-electrode with higher electron-transfer rates. Pyrolytic gas of sewage sludge is used to modify the titanium alloy. The current generation, anodic biofilm formation surface, and hydrophobicity are systematically investigated by comparing bare electrodes with three modified electrodes. Maximum current density (15.80 A/m2), achieved using a modified electrode, is 316-fold higher than that of the bare titanium alloy electrode (0.05 A/m2) and that achieved by titanium alloy electrodes modified by other methods (12.70 A/m2). The pyrolytic gas-modified titanium alloy electrode can be used as a high-performance and scalable bio-anode for bio-electrochemical systems because of its high electron-transfer rates, hydrophilic nature, and ability to achieve high current density.

  14. Effects of neutron irradiation on red blood cell labeling with technetium-99m

    International Nuclear Information System (INIS)

    Eng, R.R.; Conklin, J.J.; Grissom, M.P.

    1982-01-01

    The effects of in vivo and in vitro neutron irradiation on red blood cell radiolabeling with technetium-99m (Tc-99m) were studied. Blood from three dogs was irradiated with neutrons (725 rads, free in air dose) followed by radiolabeling with Tc-99m. The three dogs were subsequently whole body, neutron irradiated (250 rads, midline dose); and blood samples were drawn for radiolabeling at 24, 48, 72 and 96 hours post-irradiation. Blood from three control dogs was also drawn and radiolabeled on each day for comparison. The results show that there were no significant differences between the radiolabeling capacities of in vivo or in vitro neutron irradiated and control RBCs

  15. Ultrastructural analysis of metal particles released from stainless steel and titanium miniplate components in an animal model.

    Science.gov (United States)

    Matthew, I R; Frame, J W

    1998-01-01

    Low-vacuum scanning electron microscopy (Ivac SEM) was used to characterize the appearance of metal particles released from stressed and unstressed Champy miniplates placed in dogs and to study the relationship of the debris to the surrounding tissues. Under general endotracheal anesthesia, two Champy miniplates (titanium or stainless steel) were placed on the frontal bone in an animal model. One miniplate was bent to fit the curvature of the frontal bone (unstressed) and another miniplate of the same material was bent in a curve until the midpoint was raised 3 mm above the ends. The latter miniplate adapted to the skull curvature under tension during screw insertion (stressed). The miniplates and surrounding tissues were retrieved after intervals of 4, 12, and 24 weeks. Decalcified sections were prepared and examined by light microscopy and Ivac SEM. Under Ivac SEM examination, the titanium particles had a smooth, polygonal outline. Stainless steel particles were typically spherical, with numerous small projections on the surface. Most particles were 1 to 10 microns in diameter. The tissue response to the particles was variable; some particles were covered by fibrous connective tissue or enclosed by bone, and others were intracellular. The metal particles released from stressed or unstressed Champy miniplates were similar, and this was related to their source of origin and duration within the tissues. The tissue response to the particles appeared to depend on their location.

  16. Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

    Science.gov (United States)

    Takahashi, H.; Garner, F. A.

    1992-10-01

    Additions of silicon and aluminum suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminum on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminum segregates away from these boundaries. Whereas the formation of the Ni 3Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation.

  17. Microstructural evolution of neutron-irradiated Ni-Si and Ni-Al alloys

    International Nuclear Information System (INIS)

    Takahashi, H.; Garner, F.A.

    1992-01-01

    Additions of silicon and aluminium suppress the neutron-induced swelling of pure nickel but to different degrees. Silicon is much more effective initially when compared to aluminium on a per atom basis but silicon exhibits a nonmonotonic influence on swelling with increasing concentration. Silicon tends to segregate toward grain boundaries while aluminium segregates away from these boundaries. Whereas the formation of the Ni 3 Si phase is frequently observed in charged particle irradiation experiments conducted at much higher displacement rates, it did not occur during neutron irradiation in this study. Precipitation also did not occur in Ni-5Al during neutron irradiation, nor has it been reported to occur during ion irradiation. (orig.)

  18. Materials irradiation subpanel report to BESAC neutron sources and research panel

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Goland, A.N.; Lott, R.

    1992-01-01

    The future success of the nuclear power option in the US (fission and fusion) depends critically on the continued existence of a healthy national materials-irradiation program. Consideration of the requirements for acceptable materials-irradiation systems in a new neutron source has led the subcommittee to identify an advanced steady-state reactor (ANS) as a better choice than a spallation neutron source. However, the subcommittee also hastens to point out that the ANS cannot stand alone as the nation's sole high-flux mixed-spectrum neutron irradiation source in the next century. It must be incorporated in a broader program that includes other currently existing neutron irradiation facilities. Upgrading and continuing support for these facilities must be planned. In particular, serious consideration should be given to converting the HFIR into a dedicated materials test reactor, and long-term support for several university reactors should be established

  19. Efficiency determination of resistive plate chambers for fast quasi-monoenergetic neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Roeder, M.; Cowan, T.E.; Kempe, M.; Yakorev, D. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Technische Universitaet Dresden, Dresden (Germany); Elekes, Z. [MTA ATOMKI, Debrecen (Hungary); Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Aumann, T.; Caesar, C. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Technische Universitaet Darmstadt, Darmstadt (Germany); Bemmerer, D.; Sobiella, M.; Stach, D.; Wagner, A. [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany); Boretzky, K.; Hehner, J.; Heil, M. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Maroussov, V. [Universitaet zu Koeln, Koeln (Germany); GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Nusair, O. [GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt (Germany); Al-Balqa Applied University, Salt (Jordan); Prokofiev, A.V. [Uppsala University, The Svedberg Laboratory, Uppsala (Sweden); Reifarth, R. [Johann Wolfgang Goethe - Universitaet, Frankfurt am Main (Germany); Zilges, A. [Universitaet zu Koeln, Koeln (Germany); Zuber, K. [Technische Universitaet Dresden, Dresden (Germany); Collaboration: R3B Collaboration

    2014-07-15

    Composite detectors made of stainless-steel converters and multigap resistive plate chambers have been irradiated with quasi-monoenergetic neutrons with a peak energy of 175 MeV. The neutron detection efficiency has been determined using two different methods. The data are in agreement with the output of Monte Carlo simulations. The simulations are then extended to study the response of a hypothetical array made of these detectors to energetic neutrons from a radioactive ion beam experiment. (orig.)

  20. Ionic implantation by plasma in titanium and stainless steels used in prosthesis and medical instruments

    International Nuclear Information System (INIS)

    Munoz C, A. E.

    2008-01-01

    A study of a process known as plasma immersion ion implantation (PIII) of nitrogen at low voltages (< 4 kV) into three kind of samples: 1) austenitic stainless AISI 316-L steel plates, 2) ferritic stainless AISI 434 steel-based dentistry drills and 3) commercially pure titanium (CPTi) disks. On the case of CPTi the study was conducted in nitrogen- oxygen calibrated mixtures: 90% N-10% O, 80% N-20% O, 70% N-30% O and in 99.5% pure oxygen and 99.9% pure nitrogen. The PIII process was carried out by using a direct current plasma source controlled both in voltage and current, a negative voltage pulse modulator, a stainless AISI 304 steel vacuum chamber and a rod of the same material, horizontally located in the upper region of the chamber, which plays the role of anode in the plasma discharge. The purpose of the nitriding is forming a relatively thick layer on the surface of the steel specimens in order to enhance their both microhardness and general corrosion performances, desirable in medical applications. This layer contains interstitial nitrogen atoms (∼24% at.) which gives place to a deformed lattice (expanded phase) of the steel. Vickers microhardness and potentiodynamic tests (the latter in agreement to the norm ASTM G-61-89) confirm an increase of microhardness up to three times and a decrease of general corrosion rate in one order of magnitude. The nitriding of de dentistry drills is aimed at inhibiting the pitting corrosion produced by the asepsis process which results in pit nucleations, their propagation and consequent fractures when being under cyclic stress (fatigue). Scanning electron microscope micrographs reveal the risks involved in surpassing the critical treatment simple temperature of 450 C as the PIII process itself induces pitting. On its part, cyclic (ASTM G-61) potentiodynamic tests indicate an excellent pitting corrosion resistance of the samples treated under 450 C. In turn, the treatment of CPTi was meant to develop oxidized and

  1. Microchemical evolution of irradiated stainless steels

    International Nuclear Information System (INIS)

    Garner, F.A.

    1980-01-01

    The precipitates that develop during irradiation play the dominant role in the response of 300 series alloys, which alters not only the diffusional properties of point defects but also the rate of acceptance of point defects at dislocations and voids. The major elemental participants are carbon, nickel and silicon. Carbon appears to function as a major governing factor of the route and rate by which the radiation-induced evolution proceeds. It is the sensitivity of carbon's response to a wide range of variables that accounts for much of the variability observed in the swelling of 316 stainless steel. Silicon's role is two-fold: while in solution it depresses void nucleation and determines the duration of the void incubation period, and it also coprecipitates with nickel. The eventual level of nickel in the alloy matrix appears to control the steady-state swelling rate and is determined by the silicon and carbon content. The other participating elements appear to affect primarily the distribution and activity of carbon. Dislocations introduced either by irradiation or cold work likewise appear to influence the role of carbon. Several new physical mechanisms appear to be operating: Inverse Kirkendall effect, interstitial-altered phase stability, solute-interstitial binding, infiltration-exchange process, and creation of radiation-stable precipitates. The sensitivity of the latter phenomenon to temperature and flux has been shown to account for much of the unusual behavior of AISI 316 during irradiation

  2. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  3. Fast and epithermal neutron radiography using neutron irradiator

    International Nuclear Information System (INIS)

    Oliveira, Karol A.M. de; Crispim, Verginia R.; Ferreira, Francisco J.O.

    2013-01-01

    The neutron radiography technique (NR) with neutrons in the energy range fast to epithermal is a powerful tool used in no-destructive inspection of bulky objects of diverse materials, including those rich in hydrogen, oxygen, nitrogen ad carbon. Thus, it can be used to identify, inclusions, voids and thickness differences in materials such as explosive artifacts and narcotics. Aiming at using NR with fast and epithermal neutrons, an Irradiator was constructed by: a 241 Am-Be source, with 5 Ci activity, a collimator with adjustable collimation rate, L/D; and a shield device composed by plates of borated paraffin and iron. The test specimens chosen were a Beam Purity Indicator (BPI) and an Indicator of Visual Resolution (IVR). The neutron radiography images obtained had a resolution of 444.4 μm and 363.6 μm respectively when registered in: 1) the sheet of the nuclear track solid detector, CR-39 type, through X (n,p) Y nuclear reaction; and 2) Kodak Industrex M radiographic film plate in close contact with a boron converter screen, both stored in a Kodak radiographic cassette. (author)

  4. Effect of neutron irradiation on the breakdown voltage of power MOSFET's

    International Nuclear Information System (INIS)

    Hasan, S.M.Y.; Kosier, S.L.; Schrimpf, R.D.; Galloway, K.F.

    1994-01-01

    The effect of neutron irradiation on power metal-oxide-semiconductor field effect transistor (power MOSFET) breakdown voltage has been investigated. Transistors with various breakdown voltage ratings were irradiated in a TRIGA nuclear reactor with cumulative fluence levels up to 5 x 10 14 neutrons/cm 2 (1 MeV equivalent). Noticeable increases in the breakdown voltages are observed in n-type MOSFET's after 10 13 neutrons/cm 2 and in p-type MOSFETs after 10 12 neutrons/cm 2 . An increase in breakdown voltage of as much as 30% is observed after 5 x 10 14 neutrons/cm 2 . The increase in breakdown voltage is attributed to the neutron-irradiation-induced defects which decrease the mean free path and trap majority carriers in the space charge region. The effect of positive trapped oxide charge due to concomitant gamma radiation and the effect of the termination structure on the increase in breakdown voltage are considered. An empirical model is presented to predict the value of the breakdown voltage as a function of neutron fluence

  5. Behavior of fluorine 18 in neutron irradiated zeolites

    International Nuclear Information System (INIS)

    Estevez Lopez, D.R.

    1992-01-01

    The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

  6. Microstructural evolution in austenitic stainless steel irradiated with triple-beam

    Energy Technology Data Exchange (ETDEWEB)

    Hamada, Shozo; Miwa, Yukio; Yamaki, Daiju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yichuan, Zhang

    1997-03-01

    An austenitic stainless steel was simultaneously irradiated with nickel, helium and hydrogen ions at the temperature range of 573-673 K. The damage level and injected concentration of He and H ions in the triple-beam irradiated region are 57 dpa, 19000 and 18000 at.ppm, respectively. Following to irradiation, the cross sectional observation normal to the incident surface of the specimen was carried out with a transmission electron microscope. Two bands parallel to the incident surface were observed in the irradiated specimen, which consist of dislocation loops and lines of high number density. These locate in the range of the depth of 0.4 to 1.3 {mu}m and 1.8 to 2.4 {mu}m from the incident surface, respectively. The region between two bands, which corresponds to the triple beam irradiated region, shows very low number density of dislocations than that in each band. Observation with higher magnification of this region shows that fine cavities with high number density uniformly distribute in the matrix. (author)

  7. Austin: austenitic steel irradiation E 145-02 Irradiation Report

    International Nuclear Information System (INIS)

    Genet, F.; Konrad, J.

    1987-01-01

    Safety measures for nuclear reactors require that the energy which might be liberated in a reactor core during an accident should be contained within the reactor pressure vessel, even after very long irradiation periods. Hence the need to know the mechanical properties at high deformation velocity of structure materials that have received irradiation damage due to their utilization. The stainless steels used in the structures of reactors undergo damage by both thermal and fast neutrons, causing important changes in the mechanical properties of these materials. Various austenitic steels available as structural materials were irradiated or are under irradiation in various reactors in order to study the evolution of the mechanical properties at high deformation velocity as a function of the irradiation damage rate. The experiment called AUSTIN (AUstenitic STeel IrradiatioN) 02 was performed by the JRC Petten Establishment on behalf of Ispra in support of the reactor safety programme

  8. The evolution of mechanical property change in irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Lucas, G.E.

    1993-01-01

    The evolution of mechanical properties in austenitic stainless steels during irradiation is reviewed. Changes in strength, ductility and fracture toughness are strongly related to the evolution of the damage microstructure and microstructurally-based models for strengthening reasonably correlate the data. Irradiation-induced defects promote work softening and flow localization which in turn leads to significant reductions in ductility and fracture toughness beyond about 10 dpa. The effects of irradiation on fatigue appear to be modest except at high temperature where helium embrittlement becomes important. The swelling-independent component of irradiation creep strain increases linearly with dose and is relatively insensitive to material variables and irradiation temperature, except at low temperatures where accelerated creep may occur as a result of low vacancy mobility. Creep rupture life is a strong function of helium content, but is less sensitive to metallurgical conditions. Irradiation-induced stress corrosion cracking appears to be related to the evolution of radiation-induced segregation/depletion at grain boundaries, and hence may not be significant at low irradiation temperatures. (orig.)

  9. Studies of neutron irradiation effects at IPNS-REF

    International Nuclear Information System (INIS)

    Kirk, M.A.

    1983-09-01

    Neutron irradiation effects studies at the Radiation Effects Facility (REF) at the Intense Pulsed Neutron Source (IPNS) located at Argonne National Laboratory (ANL) are reviewed. A brief history of the development of this user facility is followed by an overview of the scientific program. Experiments unique to a spallation neutron source are covered in more detail. Future direction of research at this facility is suggested

  10. Characterization of hybrid self-powered neutron detector under neutron irradiation

    CERN Document Server

    Nakamichi, M; Yamamura, C; Nakazawa, M; Kawamura, H

    2000-01-01

    To evaluate the irradiation behaviour of a blanket mock-up on in-pile functional test, it is necessary to measure the neutron flux change in the in-pile mock-up by a neutron detector, such as the self-powered neutron detector (SPND). With its small-sized emitter, which has high sensitivity and fast response time, SPND is an indispensable tool in order to measure the local neutron flux change. In the case of an in-pile functional test, it is necessary that response time is less than 1s and ratio of SPND output current is more than 0.3 of output current of SPND with Rh emitter. Therefore, a hybrid SPND with high sensitivity and fast response time was developed. This hybrid SPND used a hybrid emitter, i.e. Co cladded Pt-13%Rh.

  11. Estimate of absorbed dose received by individuals irradiated with neutrons

    International Nuclear Information System (INIS)

    Fonseca, E.S. da; Mauricio, C.L.P.

    1995-01-01

    An innovating methodology is proposed to estimate the absorbed dose received by individuals irradiated with neutrons in an accident, even in the case that the victim is not using any kind of neutron dosemeter. The method combines direct measurements of 24 Na and 32 P activated in the human body. The calculation method was developed using data taken from previously published papers and experimental measurements. Other irradiations results in different neutron spectra prove the validity of the methodology here proposed. Using a whole body counter to measure 24 Na activity, it is possible to evaluate neutron absorbed doses in the order of 140 μGy of very soft (thermal) spectra. For fast neutron fields, the lower limit for neutron dose detection increases, but the present method continues to be very useful in accidents, with higher neutron doses. (author). 5 refs., 1 fig., 4 tabs

  12. Fast neutron irradiation deteriorates hippocampus-related memory ability in adult mice.

    Science.gov (United States)

    Yang, Miyoung; Kim, Hwanseong; Kim, Juhwan; Kim, Sung-Ho; Kim, Jong-Choon; Bae, Chun-Sik; Kim, Joong-Sun; Shin, Taekyun; Moon, Changjong

    2012-03-01

    Object recognition memory and contextual fear conditioning task performance in adult C57BL/6 mice exposed to cranial fast neutron irradiation (0.8 Gy) were examined to evaluate hippocampus-related behavioral dysfunction following acute exposure to relatively low doses of fast neutrons. In addition, hippocampal neurogenesis changes in adult murine brain after cranial irradiation were analyzed using the neurogenesis immunohistochemical markers Ki-67 and doublecortin (DCX). In the object recognition memory test and contextual fear conditioning, mice trained 1 and 7 days after irradiation displayed significant memory deficits compared to the sham-irradiated controls. The number of Ki-67- and DCX-positive cells decreased significantly 24 h post-irradiation. These results indicate that acute exposure of the adult mouse brain to a relatively low dose of fast neutrons interrupts hippocampal functions, including learning and memory, possibly by inhibiting neurogenesis.

  13. In-service irradiated and aged material evaluations

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Alexander, D.J.

    1995-01-01

    The objective of this task is to provide a direct assessment of actual material properties in irradiated components of nuclear reactors, including the effects of irradiation and aging. Four activities are currently in progress: (1) establishing a machining capability for contaminated or activated materials by completing procurement and installation of a computer-based milling machine in a hot cell; (2) machining and testing specimens from cladding materials removed from the Gundremmingen reactor to establish their fracture properties; (3) preparing an interpretive report on the effects of neutron irradiation on cladding; and (4) continuing the evaluation of long-term aging of austenitic structural stainless steel weld metal by metallurgically examining and testing specimens aged at 288 and 343 degrees C and reporting the results, as well as by continuing the aging of the stainless steel cladding toward a total time of 50,000 h

  14. Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Korenev, Sergey E-mail: sergey_korenev@steris.com; Sikolenko, Vadim

    2004-10-01

    The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

  15. Non-destructive diagnostics of irradiated materials using neutron scattering from pulsed neutron sources

    Science.gov (United States)

    Korenev, Sergey; Sikolenko, Vadim

    2004-09-01

    The advantage of neutron-scattering studies as compared to the standard X-ray technique is the high penetration of neutrons that allow us to study volume effects. The high resolution of instrumentation on the basis neutron scattering allows measurement of the parameters of lattice structure with high precision. We suggest the use of neutron scattering from pulsed neutron sources for analysis of materials irradiated with pulsed high current electron and ion beams. The results of preliminary tests using this method for Ni foils that have been studied by neutron diffraction at the IBR-2 (Pulsed Fast Reactor at Joint Institute for Nuclear Research) are presented.

  16. Irradiation system for neutron capture therapy using the small accelerator

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Hoshi, Masaharu

    2002-01-01

    Neutron capture therapy (NCT) is to kill tumor cells that previously incorporated the stable isotope which generates heavy charged particles with a short range and a high linear energy transfer (LET) on neutron irradiation. Boron-10 is ordinarily used as such an isotope. The tumor tissue is neutron-irradiated at craniotomy after preceding craniotomy for tumor extraction: therefore two surgeries are required for the present NCT in Japan. The reactions 10 B(n, αγ) 7 Li and 7 Li (p, n) 7 Be are thought preferential for patients and doctors if a convenient small accelerator, not the reactor used at present, is available in the hospital because only one craniotomy is sufficient. Authors' examinations of the system for NCT using the small accelerator involve irradiation conditions, desirable energy spectrum of neutron, characterization of thermal and epi-thermal neutrons, social, practical and technical comparison of the reactor and accelerator, and usefulness of the reaction 7 Li (p, n) 7 Be. The system devoted to the NCT is awaited in future. (K.H.)

  17. Morphological profiles of neutron and X-irradiated small intestine

    International Nuclear Information System (INIS)

    Carr, K.E.; O'Shea, O.; Hazzard, R.A.; McCullough, J.S.; Hume, S.P.; Nelson, A.C.

    1996-01-01

    This paper describes the response of mouse small intestine, at several time points after treatment with neutron or X-irradiation, using doses expected to give similar effects in terms of crypt/microcolony survival. Using resin histology, the effects of radiation on the numbers of duodenal cell types and measurements of tissue areas were assessed. The results for individual parameters and for an estimate of overall damage are given in a data display, which summarises the morphological profile of the organ after both types of radiation. Damage and recovery were seen for many of the parameters studied but there was no standard response pattern applicable for all parameters. In particular, the response of individual crypt cell types could not be predicted from knowledge of the change in crypt numbers. With regard to the holistic response of the gut, neutron irradiation appeared to have caused more damage and produced more early effects than the X-irradiation. More specifically, neutron treatment led to more damage to the neuromuscular components of the wall, while X-irradiation produced early vascular changes. (author)

  18. Apoptosis of nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation

    International Nuclear Information System (INIS)

    Liang Ke; He Shaoqin; Feng Yan; Tang Jinhua; Feng Qinfu; Shen Yu; Yin Weibo; Xu Guozhen; Liu Xinfan; Wang Luhua; Gao Li

    1999-01-01

    Objective: To study the apoptotic response of the nasopharyngeal carcinoma cell line (CNE-2) induced by neutron irradiation. Methods: CNE-2 cells were cultured as usual. Using the techniques of DNA agarose gel electrophoresis and DNA special fluorescent staining, the status of apoptosis in CNE-2 cells after neutron irradiation was detected. Results: It was shown that the apoptosis can be induced in CNE-2 cell after neutron radiation. Six hrs, after different doses of neutron (0/0.667/1.333/2.000/2.667/3.333 Gy) and X-ray 0/2/4/6/8/10 Gy) irradiation the apoptotic rates were 2.4%, 6.3%, 7.1%, 9.5%, 13.5%, 14.6% and 2.4%, 3.8%, 5.7%, 7.8%, 10.4%, 11.7%, respectively; at 48 hrs they were 18.3%, 21.5%, 22.8%, 29.3%, 34.2% and 13.7%, 17.6%, 21.3%, 25.6%, 28.9%, respectively. At 10 hrs after neutron irradiation the DNA ladder of apoptosis could be detected between 0.667-3.333 Gy doses in CNE-2 cells by DNA agarose gel electrophoresis. Conclusion: Neutron radiation can induce apoptosis in tumor cells. Compared with the X-ray, neutron induces apoptosis in larger extent than X-ray in the same condition; meanwhile, apoptosis after irradiation is dose and time dependent

  19. Measurement and evaluation of fast neutron flux of CT and OR5 irradiation hole in HANARO

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Choo, Kee Nam; Lee, Seung-Kyu; Kim, Yong Kyun

    2012-01-01

    The irradiation test has been conducted to evaluate the irradiation performance of many materials by a material capsule at HANARO. Since the fast neutron fluence above 1 MeV is important for the irradiation test of material, it must be measured and evaluated exactly at each irradiation hole. Therefore, a fast neutron flux was measured and evaluated by a 09M-02K capsule irradiated in an OR5 irradiation hole and a 10M-01K capsule irradiated in a CT irradiation hole. Fe, Ni, and Ti wires as the fluence monitor were used for the detection of fast neutron flux. Before the irradiation test, the neutron flux and spectrum was calculated for each irradiation hole using an MCNP code. After the irradiation test, the activity of the fluence monitor was measured by an HPGe detector and the reaction rate was calculated. For the OR5 irradiation hole, the radial difference of the fast neutron flux was observed from a calculated data due to the OR5 irradiation hole being located outside the core. Furthermore, a control absorber rod was withdrawn from the core as the increase of the irradiation time at the same irradiation cycle, so the distribution of neutron flux was changed from the beginning to the end of the cycle. These effects were considered to evaluate the fast neutron flux. Neutron spectrums of the CT and OR5 irradiation hole were adjusted by the measured data. The fluxes of a fast neutron above 1 MeV were compared with calculated and measured value. Although the maximum difference was shown at 18.48%, most of the results showed good agreement. (author)

  20. HIP bonding for the different material between Niobium and Stainless steel

    International Nuclear Information System (INIS)

    Inoue, H.; Saito, K.; Abe, K.; Fujino, T.; Hitomi, N.; Kobayashi, Y.

    2000-01-01

    In the future advanced cryomodule for superconducting RF cavities, a helium vessel made from titanium or stainless steel has to be welded directly to the niobium cavity wall in order to be simple structure. For that, we need a transformer from niobium to titanium or stainless steel. Stainless steel will have many benefits if the reliable bonding to the niobium is developed. We have tested the niobium/stainless steel bonding by HIP (Hot Isostatic Pressing) with the heat shock between 1023K and 2K. The bonding interface was also observed by SEM. These test results will be presented. (author)