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Sample records for f82h steels irradiated

  1. Microstructural study of irradiated isotopically tailored F82H steel

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    Wakai, E. E-mail: wakai@realab01.tokai.jaeri.go.jp; Miwa, Y.; Hashimoto, N.; Robertson, J.P.; Klueh, R.L.; Shiba, K.; Abiko, K.; Furuno, S.; Jitsukawa, S

    2002-12-01

    The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250-400 deg. C to 2.8-51 dpa in HFIR has been examined using isotopes of {sup 54}Fe or {sup 10}B. Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 deg. C to 2.8 dpa. Helium atoms also influenced them at around 300 deg. C, and the effect of helium atoms was enhanced at 400 deg. C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of {alpha}{sup '}-precipitates on dislocation loops.

  2. Microstructural evolution of HFIR-irradiated low activation F82H and F82H-{sup 10}B steels

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    Wakai, E.; Shiba, K.; Sawai, T. [Japan Atomic Energy Research Inst. (Japan); Hashimoto, N.; Robertson, J.P.; Klueh, R.L. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Microstructures of reduced-activation F82H (8Cr-2W-0.2V-0.04Ta) and the F82H steels doped with {sup 10}B, irradiated at 250 and 300 C to 3 and 57 dpa in the High Flux Isotope Reactor (HFIR), were examined by TEM. In the F82H irradiated at 250 C to 3 dpa, dislocation loops, small unidentified defect clusters with a high number density, and a few MC precipitates were observed in the matrix. The defect microstructure after 300 C irradiation to 57 dpa is dominated by the loops, and the number density of loops was lower than that of the F82H-{sup 10}B steel. Cavities were observed in the F82H-{sup 10}B steels, but the swelling value is insignificant. Small particles of M{sub 6}C formed on the M{sub 23}C{sub 6} carbides that were present in both steels before the irradiation at 300 C to 57 dpa. A low number density of MC precipitate particles formed in the matrix during irradiation at 300 C to 57 dpa.

  3. Investigations of void formation in neutron irradiated iron and F82H steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    2002-01-01

    In the present work pure iron and low activation steel F82H have been neutron irradiated at temperatures in the interval 50 deg.C - 350 deg.C to a dose of 0.23 dpa (displacements per atom). The formation of defects has been investigated by the use ofpositron annihilation spectroscopy (PAS......). In addition iron has been irradiated to different doses in the range 0.01 - 0.4 dpa at 50oC and 100oC and the dose dependence of the electrical conductivity determined. The results demonstrated that theformation of voids takes place during neutron irradiation of pure iron in the whole temperature range....... For irradiation temperatures of 50 deg.C and 100 deg.C also a high density of micro-voids was observed. Voids and micro-voids were also detected in lowactivation F82H steel for a low irradiation temperature (50 deg.C), while for irradiation close to the temperature of annealing stage V (250 deg.C), no voids...

  4. Effects of heat treatment and irradiation on mechanical properties in F82H steel doped with boron and nitrogen

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    Okubo, N. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan)]. E-mail: okubo.nariaki@jaea.go.jp; Wakai, E. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Matsukawa, S. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Sawai, T. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Kitazawa, S. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Shirakata 2-4, Tokai-mura, Ibaraki-ken 319-1195 (Japan)

    2007-08-01

    Effects of heat treatment and irradiation on mechanical properties and microstructures have been studied for martensitic steel F82H co-doped with 60 ppm B and 200 ppm N (F82H + B + N) to evaluate fundamental mechanical properties and irradiation response before irradiation at JMTR and HFIR facilities. The specimens were firstly normalized at 1150 {sup o}C and tempered at 700 {sup o}C, secondly normalized at 1000 {sup o}C and tempered at 700, 750 and 780 {sup o}C. The tensile properties were measured for the specimens before irradiation. Single ion irradiations of 10.5 MeV Fe{sup 3+} and dual ion irradiations of 10.5 MeV Fe{sup 3+} with simultaneous 1.05 MeV He{sup +} of 10 appmHe/dpa rate were performed at 160-590 {sup o}C to 20 dpa. Micro-hardness was measured before and after the irradiation. Tensile properties of the F82H + B + N were similar to F82H and also radiation hardening behaved similarly to F82H. The change of hardening increased with increasing temperature, saturated around 350 {sup o}C and decreased at higher temperature.

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

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    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.

  6. Effect of boron on post irradiation tensile properties of reduced activation ferritic steel (F-82H) irradiated in HFIR

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    Shiba, Kiyoyuki; Suzuki, Masahide; Hishinuma, Akimichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Dept. of Materials Science and Engineering; Pawel, J.E. [Oak Ridge National Lab., TN (United States). Metals and Ceramics Div.

    1994-12-31

    Reduced activation ferritic/martensitic steel, F-82H (Fe-8Cr-2W-V-Ta), was irradiated in the High Flux Isotope Reactor (HFIR) to doses between 11 and 34 dpa at 400 and 500 C. Post irradiation tensile tests were performed at the nominal irradiation temperature in vacuum. Some specimens included {sup 10}B or natural boron (nB) to estimate the helium effect on tensile properties. Tensile properties including the 0.2% offset yield stress, the ultimate tensile strength, the uniform elongation and the total elongation were measured. The tensile properties were not dependent on helium content in specimens irradiated to 34 dpa, however {sup 10}B-doped specimens with the highest levels of helium showed slightly higher yield strength and less ductility than boron-free specimens. Strength appears to go through a peak, and ductility through a trough at about 11 dpa. The irradiation to more than 21 dpa reduced the strength and increased the elongation to the unirradiated levels. Ferritic steels are one of the candidate alloys for nuclear fusion reactors because of their good thermophysical properties, their superior swelling resistance, and the low corrosion rate in contact with potential breeder and coolant materials.

  7. Creep behavior of the F82H steel under irradiation with 17 MeV protons at 300 deg. C

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    Nagakawa, Johsei [National Institute for Materials Science (NIMS), 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-1 Kasuga Koen, Kasuga, Fukuoka 816-8580 (Japan)], E-mail: NAGAKAWA.Johsei@nims.go.jp; Uchio, S. [National Institute for Materials Science (NIMS), 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-1 Kasuga Koen, Kasuga, Fukuoka 816-8580 (Japan); Murase, Y.; Yamamoto, N. [National Institute for Materials Science (NIMS), 1-2-1 Sengen, Tsukuba, Ibaraki 305-0047 (Japan); Shiba, K. [Japan Atomic Energy Agency (JAEA), 2-4 Shirakata, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan)

    2009-04-30

    Although fusion blankets are exposed to severe irradiation, its rear side would stay at rather a modest condition. In this research, the irradiation-induced deformation of F82H IEA-heat steel at 300 deg. C was examined. A torsion creep apparatus with a strain resolution of {approx}10{sup -7} was used with 17 MeV protons (2 x 10{sup -7} dpa/s). At the lowest stress of 30 MPa, deformation in the direction against applied stress was observed. This 'negative creep' was attributed to the increase in elastic modulus due to irradiation. Such an effect was compensated for each measurement based on the modulus data measured during irradiation. Stress exponent n of irradiation creep rates was 1.5, very close to that of creep strain at 5 dpa of pressurized tubes. The predicted stress relaxation was slower than that for 5% cold-worked Type 316L steel, resulting mainly from the difference in n, smaller and closer to unity in the latter.

  8. Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

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    Hirose, Takanori [Japan Atomic Energy Agency (JAEA); Sokolov, Mikhail A [ORNL; Ando, M. [Japan Atomic Energy Agency (JAEA); Tanigawa, H. [Japan Atomic Energy Agency (JAEA); Shiba, K. [Japan Atomic Energy Agency (JAEA); Stoller, Roger E [ORNL; Odette, G.R. [University of California, Santa Barbara

    2013-11-01

    This work investigates irradiation response in the joints of F82H employed for a fusion breeding blanket. The joints, which were prepared using welding and diffusion welding, were irradiated up to 6 dpa in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Post-irradiation tests revealed hardening in weldment (WM) and base metal (BM) greater than 300 MPa. However, the heat affected zones (HAZ) exhibit about half that of WM and BM. Therefore, neutron irradiation decreased the strength of the HAZ, leaving it in danger of local deformation in this region. Further the hardening in WM made with an electron beam was larger than that in WM made with tungsten inert gas welding. However the mechanical properties of the diffusion-welded joint were very similar to those of BM even after the irradiation.

  9. Irradiation response in weldment and HIP joint of reduced activation ferritic/martensitic steel, F82H

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T., E-mail: hirose.takanori@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Ando, M.; Tanigawa, H.; Shiba, K. [Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Stoller, R.E. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Odette, G.R. [University of California Santa Barbara, Santa Barbara, CA (United States)

    2013-11-15

    This work investigates irradiation response in the joints of F82H employed for a fusion breeding blanket. The joints, which were prepared using welding and diffusion welding, were irradiated up to 6 dpa in the High Flux Isotope Reactor at the Oak Ridge National Laboratory. Post-irradiation tests revealed hardening in weldment (WM) and base metal (BM) greater than 300 MPa. However, the heat affected zones (HAZ) exhibit about half that of WM and BM. Therefore, neutron irradiation decreased the strength of the HAZ, leaving it in danger of local deformation in this region. Further the hardening in WM made with an electron beam was larger than that in WM made with tungsten inert gas welding. However the mechanical properties of the diffusion-welded joint were very similar to those of BM even after the irradiation.

  10. Saturation behavior of irradiation hardening in F82H irradiated in the HFIR

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

    2007-07-01

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

  11. Effect of helium production on swelling of F82H irradiated in HFIR

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    Wakai, E. E-mail: wakai@realab01.tokai.jaeri.go.jp; Hashimoto, N.; Miwa, Y.; Robertson, J.P.; Klueh, R.L.; Shiba, K.; Jistukawa, S

    2000-12-01

    The effects of helium production and heat treatment on the swelling of F82H steel irradiated in the HFIR to 51 dpa have been investigated using {sup 10}B, {sup 58}Ni and {sup 60}Ni-doped specimens. The swelling of tempered F82H-std and F82H doped with {sup 10}B irradiated at 400 deg. C ranged from 0.52% to 1.2%, while the swelling of the non-tempered F82H doped with {sup 58}Ni or {sup 60}Ni was less than 0.02%. At 300 deg. C the swelling in all steels was insignificant. In the F82H + Ni, a high number of density carbides formed in the matrix at these temperatures. The production of helium atoms enhanced the swelling of the F82H steel. However, the non-tempered treatment for the F82H + Ni suppressed remarkably the swelling. The cause of low swelling in the F82H + Ni may be due to the occurrence of the high density of carbides acting as sinks or the decrease of mobility of vacancies interacting with carbon atoms in matrix.

  12. Embrittlement of irradiated F82H in the absence of irradiation hardening

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    Klueh, R.L. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)], E-mail: kluehrl@ornl.gov; Shiba, K. [Japan Atomic Energy Agency, Toki-Mura, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2009-04-30

    Neutron irradiation of 7-12% Cr ferritic/martensitic steels below 425-450 deg. C produces microstructural defects and precipitation that cause an increase in yield stress. This irradiation hardening causes embrittlement, which is observed in a Charpy impact or fracture toughness test as an increase in the ductile-brittle transition temperature. Based on observations that show little change in strength in steels irradiated above 425-450 deg. C, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study of F82H steel, significant embrittlement was observed after irradiation at 500 deg. C, but no hardening occurred. This embrittlement is apparently due to irradiation-accelerated Laves-phase precipitation. Observations of the embrittlement of F82H in the absence of irradiation hardening have been examined and analyzed with thermal-aging studies and computational thermodynamics calculations to illuminate and understand the embrittlement during irradiation.

  13. Embrittlernent of irradiated F82H in the absence of irradiation hardening

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, Ronald L [ORNL; Shiba, Kiyoyuki [ORNL; Sokolov, Mikhail A [ORNL

    2009-01-01

    Neutron irradiation of 7-12% Cr ferritic/martensitic steels below 425-450 C produces microstructural defects and precipitation that cause an increase in yield stress. This irradiation hardening causes embrittlement, which is observed in a Charpy impact or fracture toughness test as an increase in the ductile-brittle transition temperature. Based on observations that show little change in strength in steels irradiated above 425-450 C, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study of F82H steel, significant embrittlement was observed after irradiation at 500 C. This embrittlement is apparently due to irradiation-accelerated Laves-phase precipitation. Observations of the embrittlement in the absence of hardening has been examined and analyzed with thermal-aging studies and computational thermodynamics calculations to illuminate and understand the effect.

  14. Thermal aging behavior of low activation martensitic steel F82 H

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    Shiba, K.; Jitsukaw, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States)

    2007-07-01

    Full text of publication follows: Low activation ferritic/martensitic (LAF/M) steel is the primary candidate materials for a fusion reactor structural material. Characterization of F82H conducted in the IEA working group on ferritic/martensitic steels has generated many property data on this steel and many irradiation data have been also obtained. Thermal aging behavior is an important property of the structural material to be used in high-temperature environment. And it is also important to distinguish the irradiation accelerated or irradiation induced precipitation effects from irradiation results. Therefore, thermal aging behavior of F82H IEA heat was investigated in this study. Microstructure, precipitation analysis, and mechanical properties of F82H IEA heat steel (8Cr-2WVTa) after thermal aging at 450, 500, 550, 600, and 650 deg. C for 1,000, 10,000, and 30,000 hours were examined. Hardness, tensile, and Charpy impact tests were performed as mechanical property tests before and after aging. 650 deg. C aging started softening after 1000 hours and embrittlement due to Laves phase raised 70 deg. C in ductile-to-brittle transfer temperature (DBTT) and decreased upper-shelf energy (USE) to 70% of as-received material. The aging below 600 deg. C did not show much effect on the tensile properties. Especially, no change in tensile properties was observed below 500 deg. C. 550 and 600 deg. C aging showed changes in Charpy impact properties after 3000 hours, and longer aging time increases the DBTT and decreases the USE. These changes in Charpy properties were corresponding to the growth of Laves phase precipitation. (authors)

  15. Swelling of F82H irradiated at 673 K up to 51 dpa in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Miwa, Y. E-mail: miway@popsvr.tokai.jaeri.go.jp; Wakai, E.; Shiba, K.; Hashimoto, N.; Robertson, J.P.; Rowcliffe, A.F.; Hishinuma, A

    2000-12-01

    Reduced-activation ferritic/martensitic steel, F82H (8Cr-2W-0.2V-0.04Ta-0.1C), and variants doped with isotopically tailored boron were irradiated at 673 K up to 51 dpa in the high flux isotope reactor (HFIR). The concentrations of {sup 10}B in these alloys were 4, 62, and 325 appm during HFIR irradiation which resulted in the production of 4, 62 and 325 appm He, respectively. After irradiation, transmission electron microscopy (TEM) was carried out. The number density of cavities increased and the average diameter of cavities decreased with increasing amounts of {sup 10}B. The number density decreased and the average diameter increased with increasing displacement damage. Swelling increased as a function of displacement damage and He concentration.

  16. Development of an extensive database of mechanical and physical properties for reduced-activation martensitic steel F82H

    Energy Technology Data Exchange (ETDEWEB)

    Jitsukawa, S. E-mail: jitsukawa@ifmif.tokai.jaeri.go.jp; Tamura, M.; Schaaf, B. van der; Klueh, R.L.; Alamo, A.; Petersen, C.; Schirra, M.; Spaetig, P.; Odette, G.R.; Tavassoli, A.A.; Shiba, K.; Kohyama, A.; Kimura, A

    2002-12-01

    Tensile, fracture toughness, creep and fatigue properties and microstructural studies of the reduced-activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) before and after irradiation are reported. The design concept used for the development of this alloy is also introduced. A large number of collaborative test results including those generated under the International Energy Agency (IEA) implementing agreements are collected and are used to evaluate the feasibility of using reduced-activation martensitic steels for fusion reactor structural materials, with F82H as one of the reference alloys. All the specimens used in these tests were prepared from plates obtained from 5-ton heats of F82H supplied to all participating laboratories by JAERI. Many of the results have been entered into relational databases with emphasis on traceability of records on how the specimens were prepared from plates and ingots.

  17. Evaluation of grain boundary embrittlement of phosphorus added F82H steel by SSTT

    Science.gov (United States)

    Kim, Byung Jun; Kasada, Ryuta; Kimura, Akihiko; Tanigawa, Hiroyasu

    2012-02-01

    Non-hardening embrittlement (NHE) can be happened by a large amount of He on grain boundaries over 500-700 appm of bulk He without hardening at fusion reactor condition. Especially, at high irradiation temperatures (>≈420 °C), NHE accompanied by intergranular fracture affects the severe accident and the safety of fusion blanket system. Small specimen tests to evaluate fracture toughness and Charpy impact properties were carried out for F82H steels with different levels of phosphorous addition in order to simulate the effects of NHE on the shift of transition curve. It was found that the ductile to brittle transition temperature (DBTT) and reference temperature ( T0) after phosphorous addition is shifted to higher temperatures and accompanied by intergranular fracture at transition temperatures region. The master curve approach for evaluation of fracture toughness change by the degradation of grain boundary strength was carried out by referring to the ASTM E1921.

  18. Microstructural evolutions of friction stir welded F82H steel for fusion applications

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    Noh, Sang Hoon; Shim, Jae Won; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Tani Gawa, Hiro Yasu [JAEA, Rokasho (Japan); Fujii, Hideto Shi [Osaka Univ., Osaka (Japan); Kim Ura, Aki Hiko [Kyoto Univ., Kyoto (Japan)

    2012-10-15

    A blanket is the most important component functionalized as plasma confining, tritium breeding, heat exchanging, and irradiation shielding from severe thermo neutron loads in a fusion reactor. Its structure consists of first walls, side walls, a back board, and coolant channels mainly made of reduced activation ferritic/martensitic (RAFM) steel, which is the most promising candidate as a structural material for fusion reactors. To fabricate this blanket structure, some welding and joining methods have being carefully applied. However, when fusion welding, such as tungsten inert gas (TIG) welding, electron beam, and laser welding was performed between F82H and itself, the strength of welds significantly deteriorated due to the development of {delta} ferrite and precipitate dissolution. Post welding heat treatment (PWHT) should be followed to restore the initial microstructure. Nevertheless, microstructural discontinuity inevitably occurs between the weld metal, heat affected zone and base metal and this seriously degrades the entire structural stability under pulsed operation at high temperature in test blanket module (TBM). A phase transformation can also be an issue to be solved, which leads to a difficult replacement of the blanket module. Therefore, a reliable and field applicable joining technique should be developed not to accompany with PWHT after the joining process. Friction stir welding (FSW) is one of the solid state processes that does not create a molten zone at the joining area, so the degradation of the featured microstructures may be avoided or minimized. In this study, FSW was employed to join F82H steels to develop a potential joining technique for RAFM steel. The microstructural features on the joint region were investigated to evaluate the applicability of the FSW.

  19. The dose dependence of fracture toughness Of F82H steel

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States); Tanigawa, H.; Ando, M.; Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Odette, G. [UCSB, Santa-Barbara, Dept. of Mechanical Engineering UCSB, AK (United States); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States)

    2007-07-01

    Full text of publication follows: The ferritic-martensitic steel F82H is a primary candidate low-activation material for fusion applications, and it is being investigated in the joint U.S. Department of Energy-Japan Atomic Energy Agency. As a part of this program, several capsules containing fracture toughness specimens were irradiated in High-Flux Isotope Reactor. These specimens were irradiated to a wide range of doses from 3.5 to 25 dpa. The range of irradiation temperature was from 250 deg. C to 500 deg. C. This paper summarizes the changes in fracture toughness transition temperature and decrease in the ductile fracture toughness as result of various irradiation conditions. It is shown that in the 3.5 to 25 dpa dose range, irradiation temperature plays the key rote in determination of the shift of the transition temperature. Highest embrittlement observed at 250 deg.C and the lowest at 500 deg. C. At a given irradiation temperature, shift of the fracture toughness transition temperature increases slightly with dose within the studied dose range. It appears that main gain in transition temperature shift occurred during initial {approx}5 dpa of irradiation. The present data are compared to the available published trends. (authors)

  20. Underwater explosive welding of tungsten to reduced-activation ferritic steel F82H

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Daichi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Morizono, Yasuhiro [Graduate School of Science and Technology, Kumamoto University, 2-39-1 Kurokami, Kumamoto 860-8555 (Japan); Hokamoto, Kazuyuki [Institute of Pulsed Power Science, Kumamoto University, 2-39-1 Kurokami, Kumamoto 860-8555 (Japan)

    2014-10-15

    Highlights: • The underwater explosive welding was successfully applied in the joining of tungsten to F82H reduced activation ferritic steel. • Microstructure of the interface showed the formation of a wave-like interface with a thin mixed layer of tungsten and F82H. • Nanoindentation hardness results exhibited a gradual change away from the welded interface without hardened layer. • Small punch tests on the welded specimens resulted in the cracking at a center of tungsten followed by the interfacial cracking. - Abstract: The present study reports the underwater explosive welding of commercially pure tungsten onto the surface of a reduced-activation ferritic steel F82H plate. Cross-sectional observation revealed the formation of a wave-like interface, consisting of a thin mixed layer of W and F82H. The results of nanoindentation hardness testing identified a gradual progressive change in the interface, with no hardened or brittle layer being observed. Small punch tests on the welded specimens resulted in cracking at the center of the tungsten, followed by crack propagation toward both the tungsten surface and the tungsten/steel interface.

  1. Friction stir welding of F82H steel for fusion applications

    Science.gov (United States)

    Noh, Sanghoon; Ando, Masami; Tanigawa, Hiroyasu; Fujii, Hidetoshi; Kimura, Akihiko

    2016-09-01

    In the present study, friction stir welding was employed to join F82H steels and develop a potential joining technique for a reduced activation ferritic/martensitic steel. The microstructures and mechanical properties on the joint region were investigated to evaluate the applicability of friction stir welding. F82H steel sheets were successfully butt-joined with various welding parameters. In welding conditions, 100 rpm and 100 mm/min, the stirred zone represented a comparable hardness distribution with a base metal. Stirred zone induced by 100 rpm reserved uniformly distributed precipitates and very fine ferritic grains, whereas the base metal showed a typical tempered martensite with precipitates on the prior austenite grain boundary and lath boundary. Although the tensile strength was decreased at 550 °C, the stirred zone treated at 100 rpm showed comparable tensile behavior with base metal up to 500 °C. Therefore, friction stir welding is considered a potential welding method to preserve the precipitates of F82H steel.

  2. Post irradiation plastic properties of F82H derived from the instrumented tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Taguchi, T. [Neutron Science Research Center, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan)]. E-mail: taguchi@popsvr.tokai.jaeri.go.jp; Jitsukawa, S. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan); Sato, M. [KKS, JFE, Kawasaki-Ku, Kawasaki-Shi, Kanagawa-Ken 210-0855 (Japan); Matsukawa, S. [KKS, JFE, Kawasaki-Ku, Kawasaki-Shi, Kanagawa-Ken 210-0855 (Japan); Wakai, E. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan); Shiba, K. [Department of Materials Science, Japan Atomic Energy, Research Institute, Tokai-Mura, Ibaraki-Ken 319-1195 (Japan)

    2004-12-01

    F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 deg. C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 deg. C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 deg. C.

  3. Fatigue performance and cyclic softening of F82H, a ferritic martensic steel

    Energy Technology Data Exchange (ETDEWEB)

    Stubbins, J.F. [Univ. of Illinois, Urbana, IL (United States); Gelles, D.S. [Pacific Northwest National Laboratory, Richland, WA (United States)

    1996-04-01

    The room temperature fatigue performance of F82H has been examined. The fatigue life was determined in a series of strain-controlled tests where the stress level was monitored as a function of the number of accrued cycles. Fatigue lives in the range of 10{sup 3} to 10{sup 6} cycles to failure were examined. The fatigue performance was found to be controlled primarily by the elastic strain range over most of the range of fatigue lives examined. Only at low fatigue lives did the plastic strain range contribute to the response. However, when the significant plastic strain did contribute, the material showed a tendency to cyclically soften. That is the load carrying capability of the material degrades with accumulated fatigue cycles. The overall fatigue performance of the F82H alloy was found to be similiar to other advanced martensitic steels, but lower than more common low alloy steels which possess lower yield strengths.

  4. Corrosion fatigue studies on F82H mod. martensitic steel in reducing water coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Maday, M.F.; Masci, A. [ENEA, Casaccia (Italy). Centro Ricerche Energia

    1998-03-01

    Load-controlled low cycle fatigue tests have been carried out on F82H martensitic steel in 240degC oxygen-free water with and without dissolved hydrogen, in order to simulate realistic coolant boundary conditions to be approached in DEMO. It was found that water independently of its hydrogen content, determined the same fatigue life reduction compared to the base-line air results. Water cracks exhibited in their first propagation stages similar fracture morphologies which were completely missing on the air cracks, and were attributed to the action of an environment related component. Lowering frequency gave rise to an increase in F82H fatigue lifetimes without any change in cracking mode in air, and to fatigue life reduction by microvoid coalescence alone in water. The data were discussed in terms of (i) frequency dependent concurrent processes for crack initiation and (ii) frequency-dependent competitive mechanisms for crack propagation induced by cathodic hydrogen from F82H corrosion. (author)

  5. Influence of traps on the deuterium behaviour in the low activation martensitic steels F82H and Batman

    Science.gov (United States)

    Serra, E.; Perujo, A.; Benamati, G.

    1997-06-01

    A time dependent permeation method is used to measure the permeability, diffusivity and solubility of deuterium in the low activation martensitic steels F82H and Batman. The measurements cover the temperature range from 373 to 743 K which includes the onset of deuterium trapping effects on diffusivity and solubility. The results are interpreted using a trapping model. The number of trap sites and their average energies for deuterium in F82H and Batman steels are determined.

  6. Influence of traps on the deuterium behaviour in the low activation martensitic steels F82H and Batman

    Energy Technology Data Exchange (ETDEWEB)

    Serra, E. [Commission of the European Communities, Ispra (Italy). Joint Research Centre; Perujo, A. [Commission of the European Communities, Ispra (Italy). Joint Research Centre; Benamati, G. [Associazione ENEA-EURATOM sulla FUSIONE, CR Brasimone, 40032 Camungnano Bologna (Italy)

    1997-06-01

    A time dependent permeation method is used to measure the permeability, diffusivity and solubility of deuterium in the low activation martensitic steels F82H and Batman. The measurements cover the temperature range from 373 to 743 K which includes the onset of deuterium trapping effects on diffusivity and solubility. The results are interpreted using a trapping model. The number of trap sites and their average energies for deuterium in F82H and Batman steels are determined. (orig.).

  7. Application of master curve method to the evaluation of fracture toughness of F82H steels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Jun, E-mail: kim.byungjun@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1106 (Japan); Kasada, Ryuta; Kimura, Akihiko [Institute of Advanced Energy, Kyoto University, Kyoto (Japan); Wakai, Eiichi; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1106 (Japan)

    2013-11-15

    Fracture toughness data was obtained for reduced-activation ferritic (RAF) steels with different sizes of specimens (1 compact tension(CT), 1/2 CT and 1/4 CT) using the master curve (MC) method in the transition temperature region. Considering the size adjustment by ASTM E1921, effects of specimen size on the fracture toughness are not observed and the reference temperature (T{sub 0}) is around 164 K which is similar to those (154 K) of other previous studies. However, the data are not well represented by a MC, showing a rather large number of data below the lower boundary curve. Our proposed new MC was derived within the framework of the ASTM E1921 standard to apply the MC method to F82H steel. This new MC analysis can be applied to RAF steels to estimate T{sub 0} with a better description of the data scatter in the transition temperature region of fracture toughness than that of the conventional MC analysis.

  8. Creep behavior of reduced activation martensitic steel F82H injected with a large amount of helium

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, N. E-mail: yamamoto.norikazu@nims.go.jp; Murase, Y.; Nagakawa, J.; Shiba, K

    2002-12-01

    Creep response against DEMO reactor level helium was examined on F82H steel, a candidate structural material for advanced fusion systems. Helium was injected into the material at 823 K to a concentration of about 1000 appm utilizing {alpha}-particle irradiation with a cyclotron. Post-injection creep rupture tests were conducted at the same temperature. It has been demonstrated that helium brought about no significant effect on a variety of creep properties (lifetime, rupture elongation and minimum creep rate). In parallel with this, it did not cause any influence on fracture appearance. Both helium implanted and unimplanted samples were failed in a completely transcrystalline and ductile fashion. No symptom of helium induced grain boundary separation was thereby observed even after high concentration helium introduction. These facts hint a fairly good resistance of this material toward high temperature helium embrittlement even for long-time service in fusion reactors.

  9. Hardness distribution and tensile properties in an electron beam weldment of F82H irradiated in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Oka, H., E-mail: hiroshi_oka@eng.hokudai.ac.jp [Hokkaido University, Sapporo, Hokkaido (Japan); Hashimoto, N. [Hokkaido University, Sapporo, Hokkaido (Japan); Muroga, T. [National Institute for Fusion Science, Toki, Gifu (Japan); Kimura, A. [Kyoto University, Uji, Kyoto (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Yamamoto, T. [University of California Santa Barbara, Santa Barbara, CA 93106 (United States); Ohnuki, S. [Hokkaido University, Sapporo, Hokkaido (Japan)

    2014-12-15

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

  10. Creep strength and microstructure of F82H steels near tempering temperature

    Science.gov (United States)

    Shinozuka, K.; Esaka, H.; Sakasegawa, H.; Tanigawa, H.

    2015-09-01

    Creep rupture tests near the tempering temperature were performed, and the creep behavior at high temperatures and the structures of fracture specimens were investigated. Three kinds of F82H test specimens were used: IEA-heat, mod.3, and BA07. The time-to-rupture of the BA07 specimens was the longest under all the test conditions. This was because the minimum creep rates of BA07 were smallest, and a large quantity of fine precipitates of MX from the ESR treatment were considered to be effective in providing creep resistance. Although mod.3 specimens showed a high creep resistance under high stress, the time-to-rupture of mod.3 and IEA-heat were almost the same at low stress. This was because the fine tempered martensitic structure was weakened by being subjected to a high temperature for a long period. Therefore, it is considered that a large quantity of fine MX precipitates are effective for creep resistance near the tempering temperature.

  11. Creep strength and microstructure of F82H steels near tempering temperature

    Energy Technology Data Exchange (ETDEWEB)

    Shinozuka, K., E-mail: kshinozu@nda.ac.jp [National Defense Academy, Yokosuka, Kanagawa 239-8686 (Japan); Esaka, H. [National Defense Academy, Yokosuka, Kanagawa 239-8686 (Japan); Sakasegawa, H.; Tanigawa, H. [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan)

    2015-09-15

    Creep rupture tests near the tempering temperature were performed, and the creep behavior at high temperatures and the structures of fracture specimens were investigated. Three kinds of F82H test specimens were used: IEA-heat, mod.3, and BA07. The time-to-rupture of the BA07 specimens was the longest under all the test conditions. This was because the minimum creep rates of BA07 were smallest, and a large quantity of fine precipitates of MX from the ESR treatment were considered to be effective in providing creep resistance. Although mod.3 specimens showed a high creep resistance under high stress, the time-to-rupture of mod.3 and IEA-heat were almost the same at low stress. This was because the fine tempered martensitic structure was weakened by being subjected to a high temperature for a long period. Therefore, it is considered that a large quantity of fine MX precipitates are effective for creep resistance near the tempering temperature.

  12. Fracture toughness of the IEA heat of F82H ferritic/martensitic stainless steel as a function of loading mode

    Energy Technology Data Exchange (ETDEWEB)

    Li, Huaxin; Gelles, D.S. [Pacific Northwest Labs., Richland, WA (United States); Hirth, J.P. [Washington State Univ., Pullman, WA (United States)] [and others

    1997-04-01

    Mode I and mixed-mode I/III fracture toughness tests were performed for the IEA heat of the reduced activation ferritic/martensitic stainless steel F82H at ambient temperature in order to provide comparison with previous measurements on a small heat given a different heat treatment. The results showed that heat to heat variations and heat treatment had negligible consequences on Mode I fracture toughness, but behavior during mixed-mode testing showed unexpected instabilities.

  13. Dependence of precipitate formation on normalizing temperature and its impact on the heat treatment of F82H-BA07 Steel

    Energy Technology Data Exchange (ETDEWEB)

    Fukumoto, K., E-mail: fukumoto@u-fukui.ac.jp [Research Institute for Nuclear Engineering, University of Fukui, Tsuruga 914-0055 (Japan); Sakaguchi, T.; Inoue, K. [Graduate School of Nuclear Power and Energy Safety Engineering, University of Fukui, Fukui 910-8507 (Japan); Itoh, T. [Faculty of Engineering, University of Fukui, Fukui 910-8507 (Japan); Sakasegawa, H.; Tanigawa, H. [JAEA-Aomori, Rokkasho 039-3212 (Japan)

    2013-11-15

    A detailed study of the microstructural evolution during heat treatment of F82H steel at different temperatures has been carried out. By increasing the normalizing temperature from 1193 K to 1333 K, the residual austenitic and packet grain sizes increased. The chromium-enriched precipitate, M{sub 23}C{sub 6}, increased with the normalizing temperature. Tiny vanadium-rich precipitates, of the MX type, were formed on the grain boundaries when the normalizing temperature remained below 1223 K. Above 1223 K, the proportion of vanadium-enriched precipitate decreased significantly. With a two-step normalizing heat treatment, tiny tantalum- and vanadium-rich precipitates remained on the grain boundaries and within grains when the normalizing temperature of the second step remained below 1223 K. It suggested that the second-step heat treatment in a two-step normalizing heat treatment above 1223 K influenced the tantalum- and vanadium-rich MX precipitate formation in F82H steel.

  14. Tensile properties of F82H steel after aging at 400–650 °C for 100,000 h

    Energy Technology Data Exchange (ETDEWEB)

    Nagasaka, Takuya, E-mail: nagasaka@nifs.ac.jp [National Institute for Fusion Science, Toki (Japan); Sakasegawa, Hideo; Tanigawa, Hiroyasu; Ando, Masami [Japan Atomic Energy Agency, Rokkasho (Japan); Tanaka, Teruya; Muroga, Takeo; Sagara, Akio [National Institute for Fusion Science, Toki (Japan)

    2015-10-15

    Highlights: • The present study investigated tensile properties of F82H steel after aging at 400–650 °C for 100,000 h. • Although tensile strength was degraded during the aging, the degradation of yield strength and ultimate tensile strength was 50 MPa or less at 550 °C and below in both room temperature and high temperature tests. • Above 550 °C, yield strength and ultimate tensile strength decreased with increasing aging temperature. • On the other hand, no degradation of ductility was detected after the aging. • The mechanisms for the degradation of tensile strength due to the long-term aging are discussed with previous results obtained by extraction residue analyses. - Abstract: The present study investigated tensile properties of F82H steel after aging at 400–650 °C for 100,000 h (100 kh), and discusses the mechanisms for change in tensile properties due to the long-term aging. Tensile strength was degraded during the aging. The degradation of yield strength and ultimate tensile strength was 50 MPa or less at 550 °C and below in both room temperature and high temperature tests. Above 550 °C, yield strength and ultimate tensile strength decreased with increasing aging temperature. On the other hand, no degradation of ductility was detected after the aging. Most of the degradation of the tensile strength could be attributed to loss of solid solution hardening by W due to precipitation of Laves phase (Fe{sub 2}W) at 550 °C. However, other mechanisms, such as coarsening of martensite structure and recovery of dislocations, should be taken into account to explain additional degradation above 550 °C.

  15. Macro and microscale mechanical testing and local electrode atom probe measurements of STIP irradiated F82H, Fe-8Cr ODS and Fe-8Cr-2W ODS

    Energy Technology Data Exchange (ETDEWEB)

    Hosemann, P., E-mail: peterh@lanl.gov [Los Alamos National Laboratory (LANL), MST-8 (United States); University of California Berkeley, Department of Nuclear Engineering (United States); Stergar, E. [University of California Berkeley, Department of Nuclear Engineering (United States); Peng, L. [Paul Scherrer Institute (PSI), 5332 Villigen PSI (Switzerland); Institute of Plasma Physics, Chinese Academy of Science (China); Dai, Y. [Paul Scherrer Institute (PSI), 5332 Villigen PSI (Switzerland); Maloy, S.A. [Los Alamos National Laboratory (LANL), MST-8 (United States); Pouchon, M.A. [Paul Scherrer Institute (PSI), 5332 Villigen PSI (Switzerland); Shiba, K.; Hamaguchi, D. [Japan Atomic Energy Agency (JAEA) (Japan); Leitner, H. [MontanuniversitaetLeoben, Department fuerMetallkunde (Austria)

    2011-10-01

    The reduced activation ferritic/martensitic alloy F82H (Fe-8Cr-2W-0.2V-0.04Ta-0.1C) is being considered as a structural material for several different fusion related nuclear applications. The oxide dispersion strengthened (ODS) alloys Fe-8Cr-2W ODS and Fe-8Cr ODS were developed for better high-temperature strength and radiation tolerance. These materials have been exposed to a neutron and proton environment in the Spallation Target Irradiation Program (STIP) (<13 dpa) with an average He/dpa ratio of 60 appm He/dpa at irradiation temperatures 159-347 deg. C. After irradiation, the samples were tensile tested at different temperatures. The post tensile testing fractured parts were collected and nanoindentation, microcompression testing and local electrode atom probe was conducted. The information gained by local electron atom probe in combination with the micro, nano and macroscopic mechanical tests allows one to establish a fundamental understanding of the relationship between the data measured at different scales on irradiated materials.

  16. Macro and microscale mechanical testing and local electrode atom probe measurements of STIP irradiated F82H, Fe-8Cr ODS and Fe-8Cr-2W ODS

    Science.gov (United States)

    Hosemann, P.; Stergar, E.; Peng, L.; Dai, Y.; Maloy, S. A.; Pouchon, M. A.; Shiba, K.; Hamaguchi, D.; Leitner, H.

    2011-10-01

    The reduced activation ferritic/martensitic alloy F82H (Fe-8Cr-2W-0.2V-0.04Ta-0.1C) is being considered as a structural material for several different fusion related nuclear applications. The oxide dispersion strengthened (ODS) alloys Fe-8Cr-2W ODS and Fe-8Cr ODS were developed for better high-temperature strength and radiation tolerance. These materials have been exposed to a neutron and proton environment in the Spallation Target Irradiation Program (STIP) (<13 dpa) with an average He/dpa ratio of 60 appm He/dpa at irradiation temperatures 159-347 °C. After irradiation, the samples were tensile tested at different temperatures. The post tensile testing fractured parts were collected and nanoindentation, microcompression testing and local electrode atom probe was conducted. The information gained by local electron atom probe in combination with the micro, nano and macroscopic mechanical tests allows one to establish a fundamental understanding of the relationship between the data measured at different scales on irradiated materials.

  17. Preliminary results of the round-robin testing of F82H

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K.; Yamanouchi, N.; Tohyama, A.

    1996-10-01

    Preliminary results of metallurgical, physical and mechanical properties of low activation ferritic steel F82H (IEA heat) were obtained in the round-robin test in Japan. The properties of IEA heat F82H were almost the same as the original F82H.

  18. Corrosion rate of parent and weld materials of F82H and JPCA steels under LBE flow with active oxygen control at 450 and 500 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Kenji [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)], E-mail: kikuchi.kenji21@jaea.go.jp; Kamata, Kinya; Ono, Mikinori; Kitano, Teruaki; Hayashi, Kenichi [Mitsui Engineering and Ship-building Co., Ltd., 5-6-4 Tsukiji, Chuo-ku, Tokyo 104-8439 (Japan); Oigawa, Hiroyuki [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2008-06-30

    Corrosion behavior of parent and weld materials of F82H and JPCA was studied in the circulating LBE loop under impinging flow. These are candidate materials for Japanese Accelerator Driven System (ADS) beam windows. Maximum temperatures were kept to 450 and 500 deg. C with 100 deg. C constant temperature difference. Main flow velocity was 0.4-0.6 m/s in every case. Oxygen concentration was controlled to 2-4 x 10{sup -5} mass% although there was one exception. Testing time durations were 500-3000 h. Round bar type specimens were put in the circular tube of the loop. An electron beam weld in the middle of specimens was also studied. Optical microscopy, electron microscopy, X-ray element analyses and X-ray diffraction were used to investigate corrosion in these materials. Consequently corrosion depth and stability of those oxide layers were characterized based on the analyses. For a long-term behavior a linear law is recommended to predict corrosion in the ADS target design.

  19. Tensile behavior of F82H with and without spectral tailoring

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K. E-mail: shiba@realab01.tokai.jaeri.go.jp; Klueh, R.L.; Miwa, Y.; Robertson, J.P.; Hishinuma, A

    2000-12-01

    The effects of neutron spectrum on tensile properties of the low-activation martensitic steel F82H (8Cr-2WVTa) was examined using a thermal neutron shield to tailor the neutron spectrum for steels irradiated in the high flux isotope reactor (HFIR). The yield stresses of spectrally tailored specimens irradiated in HFIR to 5 dpa at 300 deg. C and 500 deg. C are on trend lines obtained from unshielded irradiation in HFIR. No significant effect of the neutron spectrum on tensile properties could be detected.

  20. Corrosion behavior of F82H exposed to high temperature pressurized water with a rotating apparatus

    Science.gov (United States)

    Kanai, A.; Kasada, R.; Nakajima, M.; Hirose, T.; Tanigawa, H.; Enoeda, M.; Konishi, S.

    2014-12-01

    The present study reports the corrosion behavior of a reduced-activation ferritic martensitic steel F82H exposed to high temperature pressurized water for 28 and 100 h using a rotating disk apparatus at rotation speeds of 500 and 1000 rpm at a temperature of 573 K under a water pressure of 15 MPa with corrosion and/or flow-accelerated corrosion of F82H under the rotating condition.

  1. Embrittlement of reduced-activation ferritic/martensitic steels irradiated in HFIR at 300 deg. C and 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. E-mail: ku2@ornl.gov; Sokolov, M.A.; Shiba, K.; Miwa, Y.; Robertson, J.P

    2000-12-01

    Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300 deg. C and 400 deg. C in the high flux isotope reactor (HFIR) to a maximum dose of {approx}12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr-2WVTa, and 9Cr-2WVTa-2Ni, the 9Cr-2WVTa containing 2% Ni to produce helium by (n,{alpha}) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile-brittle transition temperature ({delta}DBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar {delta}DBTTs after irradiation at 300 deg. C, but after irradiation at 400 deg. C, the {delta}DBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr-2WVTa steel irradiated at 400 deg. C than for the two F82H steels. The 9Cr-2WVTa-2Ni steel with {approx}115 appm He had a larger {delta}DBTT than the 9Cr-2WVTa with {approx}5 appm He, indicating a possible helium effect.

  2. Hydrogen permeation behavior through F82H at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Matsuda, S.; Katayama, K.; Shimozori, M.; Fukada, S. [Interdisciplinary Graduate School of Engineering Science, Kyushu University, Kyushu (Japan); Ushida, H. [Energy Science and Engineering, Faculty of Engineering, Kyushu University, Kyushu (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur (Malaysia)

    2015-03-15

    F82H is a primary candidate of structural material and coolant pipe material in a blanket of a fusion reactor. Understanding tritium permeation behavior through F82H is important. In a normal operation of a fusion reactor, the temperature of F82H will be controlled below 550 C. degrees because it is considered that F82H can be used up to 30,000 hours at 550 C. degrees. However, it is necessary to assume the situation where F82H is heated over 550 C. degrees in a severe accident. In this study, hydrogen permeation behavior through F82H was investigated in the temperature range from 500 to 800 C. degrees. In some cases, water vapor was added in a sample gas to investigate an effect of water vapor on hydrogen permeation. The permeability of hydrogen in the temperature range from 500 to 700 C. degrees agreed well with the permeability reported by E. Serra et al. The degradation of the permeability by water vapor was not observed. After the hydrogen permeation reached in a steady state at 700 C. degrees, the F82H sample was heated to 800 C. degrees. The permeability of hydrogen through F82H sample which was once heated up to 800 C. degrees was lower than that of the original one. (authors)

  3. Material properties of the F82H melted in an electric arc furnace

    Energy Technology Data Exchange (ETDEWEB)

    Sakasegawa, Hideo, E-mail: sakasegawa.hideo@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Kano, Sho; Abe, Hiroaki [Institute for Materials Research, Tohoku university, Sendai, Miyagi (Japan)

    2015-10-15

    Highlights: • We studied material properties of reduced activation ferritic/martensitic steel. • We melted F82H using a 20 tons electric arc furnace for the first time. • Mass effect likely affected material properties. • MX (M: Metal, C: Carbon and/or Nitrogen) precipitates mainly formed on grain and sub grain boundaries. - Abstract: Fusion DEMO reactor requires over 11,000 tons of reduced activation ferritic/martensitic steel. It is necessary to develop the manufacturing technology for fabricating such large-scale steel with appropriate mechanical properties. In this work, we focused fundamental mechanical properties and microstructures of F82H-BA12 heat which was melted using a 20 tons electric arc furnace followed by electroslag remelting process. Its raw material of iron was blast furnace iron, because the production volume of electrolytic iron which has been used in former heats, is limited. After melting and forging, this F82H-BA12 heat was heat-treated in four different conditions to consider their fluctuations and to optimize them, and tensile and Charpy impact tests were then performed. The result of these mechanical properties were comparable to those of former F82H heats less than 5 tons which were melted applying vacuum induction melting.

  4. Effects of helium implantation on fatigue properties of F82H-IEA heat

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, N.; Murase, Y.; Nagakawa, J. [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)

    2007-07-01

    Full text of publication follows: Ferritic steels including reduced activation ones that have been recognized as attractive structural candidates for DEMO reactors and the beyond are known to be highly resistant to helium embrittlement. However, almost studies that deduced this behavior have been carried out by means of short time experiments such as tensile tests, and a few results are available concerning long term inspections, although the detrimental helium effect appears more severely in the latter. The aim of this work is to obtain further information on the influence of helium on fatigue properties of a representative reduced activation ferritic/martensitic steel F82H (8Cr2WVTa) using helium implantation technique with a cyclotron. The material examined is an IEA heat version of F82H. In order to realize a fine grain size due to thin specimens (0.08 mm thick) for ion irradiation, normalizing was conducted at rather low temperature of 1213 K, followed by tempering at 1023 K. Helium was implanted by {alpha}-particle irradiation at 823 K, a desired highest temperature of this material for first wall application, to the concentration of 100 appm He with an implantation rate of about 1.7 x 10{sup -3} appm He/s. Subsequent fatigue tests were conducted at the same temperature as that of irradiation, not only on implanted specimens but also on reference controls which were not implanted with helium but experienced the same metallurgical histories as those of irradiated ones. After fracture, samples were observed with electron microscopes. In short time periods, it has been notified that helium introduction caused no significant deterioration of both fatigue life and extension at fracture. In addition, all specimens failed in a fully trans-crystalline and ductile manner, irrespective of whether helium was present or not. Indication of grain boundary embrittlement was therefore not discerned. These facts would reflect insusceptible characteristics of this material to

  5. Mechanical properties of TIG and EB weld joints of F82H

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, Takanori, E-mail: hirose.takanori@jaea.go.jp; Sakasegawa, Hideo; Nakajima, Motoki; Tanigawa, Hiroyasu

    2015-10-15

    Highlights: • Narrow groove TIG minimized volume of F82H weld. • Mechanical properties of TIG and EB welds of F82H have been characterized. • Post weld heat treatment successfully moderate the toughness of weld metal without softening the base metal. - Abstract: This work investigates mechanical properties of weld joints of a reduced activation ferritic/martensitic steel, F82H and effects of post weld heat treatment on the welds. Vickers hardness, tensile and Charpy impact tests were conducted on F82H weld joints prepared using tungsten-inert-gas and electron beam after various heat treatments. Although narrow groove tungsten-inert-gas welding reduced volume of weld bead, significant embrittlement was observed in a heat affected zone transformed due to heat input. Post weld heat treatment above 993 K successfully moderated the brittle transformed region. The hardness of the brittle region strongly depends on the heat treatment temperature. Meanwhile, strength of base metal was slightly reduced by the treatment at temperature ranging from 993 to 1053 K. Moreover, softening due to double welding was observed only in the weld metal, but negligible in base metal.

  6. Effects of heat treatment process for blanket fabrication on mechanical properties of F82H

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. E-mail: hiroset@fusion.naka.jaeri.go.jp; Shiba, K.; Sawai, T.; Jitsukawa, S.; Akiba, M

    2004-08-01

    The objectives of this work are to evaluate the effects of thermal history corresponding to a blanket fabrication process on Reduced Activation Ferritic/Martensitic steel (RAF/Ms) microstructure, and to establish appropriate Hot Isostatic Pressing (HIP) conditions without degradation in the microstructures. One of RAF/Ms F82H and its modified versions were investigated by metallurgical methods after isochronal heat treatments up to 1473 K simulating HIP thermal history. Although conventional F82H showed significant grain growth after conventional solid HIP conditions, F82H with 0.1 wt% tantalum maintained a fine grain structure after the same heat treatment. It is considered that the grain coarsening was caused by dissolution of tantalum-carbide which immobilizes grain boundaries. On the other hands, conventional RAF/Ms with coarse grains were recovered by post HIP normalizing at temperatures below the TaC solvus temperature. This process can refine the grain size of F82H to more than ASTM grain size number 7.

  7. Joining of 14YWT and F82H by friction stir welding

    Energy Technology Data Exchange (ETDEWEB)

    Hoelzer, D.T., E-mail: hoelzerd@ornl.gov; Unocic, K.A.; Sokolov, M.A.; Feng, Z.

    2013-11-15

    Friction stir welding was investigated for joining specimens of the ODS 14YWT ferritic alloy together and to an F82H tempered martensitic steel plate. The FSW run was performed using a polycrystalline boron nitride tool and resulted in good bonding between 14YWT/14YWT and 14YWT/F82H. Joints and interfaces were observed by light microscopy and SEM analysis to be narrow in width. The ultra-small grain size of 14YWT increased by a factor up to 4 while that of F82H decreased by a considerable amount in the weld zones. The TEM analysis showed no significant changes in the size of the oxygen-enriched nanoclusters in the weld zone of 14YWT. However, defects such as a wormhole on the advancing side of the weld zone in 14YWT and small pores associated with joints and interfaces were observed in the FSW sample. The hardness measurements from unaffected zone into weld zones showed ∼20% decrease in hardness for 14YWT (from ∼500 VH to ∼380 VH) and ∼100% increase in hardness of F82H (from ∼220 VH to ∼440 VH)

  8. Determination of hydrogen diffusion coefficients in F82H by hydrogen depth profiling with a tritium imaging plate technique

    Energy Technology Data Exchange (ETDEWEB)

    Higaki, M.; Otsuka, T.; Hashizume, K. [Interdisciplinary Graduate School of Engineering and Sciences, Kyushu University, Kasuga, Fukuoka (Japan); Tokunaga, K. [Research Institute of Applied Mechanics, Kyushu University, Kasuga, Fukuoka (Japan); Ezato, K.; Suzuki, S.; Enoeda, M.; Akiba, M. [Japan Atomic Energy Agency - JAEA, Naka, Ibaraki (Japan)

    2015-03-15

    Hydrogen diffusion coefficients in a reduced activation ferritic/martensitic steel (F82H) and an oxide dispersion strengthened F82H (ODS-F82H) have been determined from depth profiles of plasma-loaded hydrogen with a tritium imaging plate technique (TIPT) in the temperature range from 298 K to 523 K. Data on hydrogen diffusion coefficients, D, in F82H, are summarized as D [m{sup 2}*s{sup -1}] =1.1*10{sup -7}exp(-16[kJ mol{sup -1}]/RT). The present data indicate almost no trapping effect on hydrogen diffusion due to an excess entry of energetic hydrogen by the plasma loading, which results in saturation of the trapping sites at the surface and even in the bulk. In the case of ODS-F82H, data of hydrogen diffusion coefficients are summarized as D [m{sup 2}*s{sup -1}] =2.2*10{sup -7}exp(-30[kJ mol{sup -1}]/RT) indicating a remarkable trapping effect on hydrogen diffusion caused by tiny oxide particles (Y{sub 2}O{sub 3}) in the bulk of F82H. Such oxide particles introduced in the bulk may play an effective role not only on enhancement of mechanical strength but also on suppression of hydrogen penetration by plasma loading.

  9. Destructive and non-destructive evaluation methods of interface on F82H HIPed joints

    Energy Technology Data Exchange (ETDEWEB)

    Kishimoto, Hirotatsu, E-mail: hkishi@mmm.muroran-it.ac.jp [OASIS, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Graduate School, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Muramatsu, Yusuke [Graduate School, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Asakura, Yuki [OASIS, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Graduate School, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Endo, Tetsuo [Graduate School, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan); Kohyama, Akira [OASIS, Muroran Institute of Technology, 27-1, Muroran, Hokkaido (Japan)

    2016-11-01

    Highlights: • The first wall of F82H steel will be fabricated by the HIP method. • Inspection techniques need to be developed for the HIPed interface. • Both destructive and non-destructive inspection techniques are introduced. - Abstract: The first walls of F82H steel with built-in cooling channels will be assembled thin plates and rectangular pipes by a HIP method. Silicon oxides form on an interface of HIPed joints during HIPing and result in the lowering of toughness of the HIPed joints. A large issue is investigation method of HIPed interface. The flexibility of specimen size for the investigation will be necessary because of the thin wall of cooling channels. A small specimen destructive test technique which is able to distinguish a base metal and an excellent HIPed joint has been desired, and recent researches find out a torsion test method to solve the issue. Non-destructive test technique is another issue for the inspection of the first wall. An ultrasonic inspection method is a candidate but silicon oxides are too small to produce good flaw echo from oxides, some solutions will be necessary. Present research introduces the current status of development of small specimen destructive test technique and the ultrasonic method for the first wall inspection.

  10. Plastic flow properties and fracture toughness characterization of unirradiated and irradiated tempered martensitic steels

    Science.gov (United States)

    Spätig, P.; Bonadé, R.; Odette, G. R.; Rensman, J. W.; Campitelli, E. N.; Mueller, P.

    2007-08-01

    We investigate the plastic flow properties at low and high temperature of the tempered martensitic steel Eurofer97. We show that below room temperature, where the Peierls friction on the screw dislocation is active, it is necessary to modify the usual Taylor's equation between the flow stress and the square root of the dislocation density and to include explicitly the Peierls friction stress in the equation. Then, we compare the fracture properties of the Eurofer97 with those of the F82H steel. A clear difference of the fracture toughness-temperature behavior was found in the low transition region. The results indicate a sharper transition for Eurofer97 than for the F82H. Finally, the shift of the median toughness-temperature curve of the F82H steel was determined after two neutron irradiations performed in the High Flux Reactor in Petten.

  11. Micro-structure and micro-hardness of ODS steels after ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Liu, C., E-mail: liuchxin@eng.hokudai.ac.jp [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Yu, C.; Hashimoto, N.; Ohnuki, S. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Ando, M.; Shiba, K.; Jitsukawa, S. [Tokai Laboratory, JAEA, Tokai, Ibaraki 319-1195 (Japan)

    2011-10-01

    The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5 MeV Fe{sup 3+} ions up to a dose of 20 dpa at 250 and 380 deg. C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation.

  12. Micro-structure and micro-hardness of ODS steels after ion irradiation

    Science.gov (United States)

    Liu, C.; Yu, C.; Hashimoto, N.; Ohnuki, S.; Ando, M.; Shiba, K.; Jitsukawa, S.

    2011-10-01

    The radiation-hardening of oxide dispersion strengthened (ODS) alloys was examined using ion irradiation and nano-indentation. In this work, three ODS steels were irradiated in the TIARA facility at JAEA with 10.5 MeV Fe 3+ ions up to a dose of 20 dpa at 250 and 380 °C. Micro-hardness measurements were carried out on the ion-irradiated specimens with ultra-low load indention. Micro-structures were investigated by transmission electron microscopy (TEM) to examine the contribution of various types of defects to the radiation-hardening. All three steels showed increases in the hardness after the ion irradiation, and F82H-ODS showed the lowest radiation-hardening, which suggests that F82H-ODS has the better radiation resistance. Small amounts of particle dissolution was also confirmed in all of the steels after the irradiation.

  13. Creep behavior of reduced activation ferritic/martensitic steels irradiated at 573 and 773 K up to 5 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Ando, M. [Fusion Structural Materials Development Group, Directorates of Fusion Energy Technology, Fusion Research and Development Directorate, Japan Atomic Energy Agency (JAEA) (Japan)]. E-mail: ando.masami@jaea.go.jp; Li, M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Tanigawa, H. [Fusion Structural Materials Development Group, Directorates of Fusion Energy Technology, Fusion Research and Development Directorate, Japan Atomic Energy Agency (JAEA) (Japan); Grossbeck, M.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996-2300 (United States); Kim, S. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Sawai, T. [Fusion Structural Materials Development Group, Directorates of Fusion Energy Technology, Fusion Research and Development Directorate, Japan Atomic Energy Agency (JAEA) (Japan); Shiba, K. [Fusion Structural Materials Development Group, Directorates of Fusion Energy Technology, Fusion Research and Development Directorate, Japan Atomic Energy Agency (JAEA) (Japan); Kohno, Y. [Muroran Institute of Technology, Muroran, Hokkaido 050-8585 (Japan); Kohyama, A. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2007-08-01

    The irradiation creep behavior of F82H and several variants of JLF-1 steel has been measured at 573 and 773 K up to 5 dpa using helium-pressurized creep tubes irradiated in HFIR. These tubes were pressurized with helium to hoop stress levels of 0-400 MPa at the irradiation temperature. The results for F82H and JLF-1 with a 400 MPa hoop stress showed small creep strains (<0.25%) after irradiation at 573 K. The irradiation creep strain at 573 K in these steels is linearly dependent on the applied stress at stress levels below 250 MPa. However, at higher hoop stress levels, the creep strain becomes nonlinear. At 773 K, the irradiation creep strain of F82H is linearly dependent on the applied stress level below 100 MPa. At higher stress levels, the creep strain increased strongly. The creep compliance coefficients for F82H and JLF-1 are consistent with the values obtained for other steels. These data contribute to the materials database for ITER test blanket design work.

  14. Embrittlement behavior of neutron irradiated RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 {sup o}C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 {sup o}C and 300 {sup o}C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 {sup o}C and 350 {sup o}C.

  15. Embrittlement behavior of neutron irradiated RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2007-08-01

    The effects of neutron irradiation on the embrittlement behavior of reduced activation ferritic/martensitic (RAFM) steel EUROFER97 for different heat treatment conditions have been investigated. The irradiation to 16.3 dpa at different irradiation temperatures (250-450 °C) was carried out in the Petten High Flux Reactor in the framework of the HFR Phase-IIb (SPICE) irradiation project. Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X) and MANET-I were also irradiated at selected temperatures. The embrittlement behavior and hardening were investigated by instrumented Charpy-V tests with subsize specimens. The neutron irradiation induced embrittlement and hardening of as-delivered EUROFER97 are comparable to those of investigated reference steels, being mostly pronounced for 250 °C and 300 °C irradiation temperatures. Heat treatment of EUROFER97 at higher austenization temperature substantially improves the embrittlement behavior at irradiation temperatures of 250 °C and 350 °C.

  16. Evaluation of Cu as an interlayer in Be/F82H diffusion bonds for ITER TBM

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, R.M., E-mail: rhunt@ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, 44-128 Engineering IV, 420 Westwood Plaza, Los Angeles, CA 90025-1597 (United States); Goods, S.H., E-mail: shgoods@sandia.gov [Sandia National Laboratories, 7011 East Ave., Livermore, CA 94550 (United States); Ying, A., E-mail: ying@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, 44-128 Engineering IV, 420 Westwood Plaza, Los Angeles, CA 90025-1597 (United States); Dorn, C.K., E-mail: christopher_dorn@brushwellman.com [Brush Wellman Inc., 14710 W. Portage River So. Road, Elmore, OH 43416 (United States); Abdou, M., E-mail: abdou@fusion.ucla.edu [Mechanical and Aerospace Engineering Department, UCLA, 44-128 Engineering IV, 420 Westwood Plaza, Los Angeles, CA 90025-1597 (United States)

    2011-10-01

    Copper has been investigated as a potential interlayer material for diffusion bonds between beryllium and Reduced Activation Ferritic/Martensitic (RAFM) steel. Utilizing Hot Isostatic Pressing (HIP), copper was directly bonded to a RAFM steel, F82H, at 650 deg. C, 700 deg. C, 750 deg. C, 800 deg. C and 850 deg. C, under 103 MPa for 2 h. Interdiffusion across the bonded interface was limited to 1 {mu}m or less, even at the highest HIP'ing temperature. Through mechanical testing it was found that samples HIP'ed at 750 deg. C and above remain bonded up to 211 MPa under tensile loading, at which point ductile failure occurred in the bulk copper. As titanium will be used as a barrier layer to prevent the formation of brittle Be/Cu intermetallics, additional annealing studies were performed on copper samples coated with a titanium thin film to study Ti/Cu interdiffusion characteristics. Samples were heated to temperatures between 650 deg. C and 850 deg. C for 2 h in order to mimic the range of likely HIP temperatures. A correlation was drawn between HIP temperature and diffusion depth for use in determining the minimum Ti film thickness necessary to block diffusion in the Be/F82H joint.

  17. Evaluation of Cu as an interlayer in Be/F82H diffusion bonds for ITER TBM

    Science.gov (United States)

    Hunt, R. M.; Goods, S. H.; Ying, A.; Dorn, C. K.; Abdou, M.

    2011-10-01

    Copper has been investigated as a potential interlayer material for diffusion bonds between beryllium and Reduced Activation Ferritic/Martensitic (RAFM) steel. Utilizing Hot Isostatic Pressing (HIP), copper was directly bonded to a RAFM steel, F82H, at 650 °C, 700 °C, 750 °C, 800 °C and 850 °C, under 103 MPa for 2 h. Interdiffusion across the bonded interface was limited to 1 μm or less, even at the highest HIP'ing temperature. Through mechanical testing it was found that samples HIP'ed at 750 °C and above remain bonded up to 211 MPa under tensile loading, at which point ductile failure occurred in the bulk copper. As titanium will be used as a barrier layer to prevent the formation of brittle Be/Cu intermetallics, additional annealing studies were performed on copper samples coated with a titanium thin film to study Ti/Cu interdiffusion characteristics. Samples were heated to temperatures between 650 °C and 850 °C for 2 h in order to mimic the range of likely HIP temperatures. A correlation was drawn between HIP temperature and diffusion depth for use in determining the minimum Ti film thickness necessary to block diffusion in the Be/F82H joint.

  18. Reduced activation martensitic steels as a structural material for ITER test blanket

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, K. E-mail: shiba@realab01.tokai.jaeri.go.jp; Enoeda, M.; Jitsukawa, S

    2004-08-01

    A Japanese ITER test blanket module (TBM) is planed to use reduced-activation martensitic steel F82H. Feasibility of F82H for ITER test blanket module is discussed in this paper. Several kinds of property data, including physical properties, magnetic properties, mechanical properties and neutron-irradiation data on F82H have been obtained, and these data are complied into a database to be used for the designing of the ITER TBM. Currently obtained data suggests F82H will not have serious problems for ITER TBM. Optimization of F82H improves the induced activity, toughness and HIP resistance. Furthermore, modified F82H is resistant to temperature instability during material production.

  19. High-dose neutron irradiation embrittlement of RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: ermile.gaganidze@imf.fzk.de; Schneider, H.-C. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Dafferner, B. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany); Aktaa, J. [Forschungszentrum Karlsruhe, Institut fuer Materialforschung II, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 deg. C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = {delta}DBTT/{delta}{sigma} indicates hardening-dominated embrittlement at irradiation temperatures below 350 deg. C with 0.17 {<=} C {<=} 0.53 deg. C/MPa. Scattering of C at irradiation temperatures above 400 deg. C indicates no hardening embrittlement.

  20. High-dose neutron irradiation embrittlement of RAFM steels

    Science.gov (United States)

    Gaganidze, E.; Schneider, H.-C.; Dafferner, B.; Aktaa, J.

    2006-09-01

    Neutron irradiation-induced embrittlement of the reduced-activation ferritic/martensitic (RAFM) steel EUROFER97 was studied under different heat treatment conditions. Irradiation was performed in the Petten High Flux Reactor within the HFR Phase-IIb (SPICE) irradiation project up to 16.3 dpa and at different irradiation temperatures (250-450 °C). Several reference RAFM steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) were also irradiated at selected temperatures. The impact properties were investigated by instrumented Charpy-V tests with subsize specimens. Embrittlement and hardening of as-delivered EUROFER97 steel are comparable to those of reference steels. Heat treatment of EUROFER97 at a higher austenitizing temperature substantially improves the embrittlement behaviour at low irradiation temperatures. Analysis of embrittlement in terms of the parameter C = ΔDBTT/Δ σ indicates hardening-dominated embrittlement at irradiation temperatures below 350 °C with 0.17 ⩽ C ⩽ 0.53 °C/MPa. Scattering of C at irradiation temperatures above 400 °C indicates no hardening embrittlement.

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, Materials Science and Technology Division, P.O. 2008 MS6138, Oak Ridge, TN 37831-6138 (United States)], E-mail: kluehrl@ornl.gov; Shiba, K. [Japan Atomic Energy Agency, Toki-Mura, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Materials Science and Technology Division, P.O. 2008 MS6138, Oak Ridge, TN 37831-6138 (United States)

    2008-07-15

    Irradiation damage caused by neutron irradiation below 425-450 deg. C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 deg. C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 deg. C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.

  2. Mechanical properties and microstructure of F-82H welded joints using CO{sub 2} laser beam

    Energy Technology Data Exchange (ETDEWEB)

    Yamanouchi, N.; Shiba, K.

    1996-10-01

    The laser welding of F-82H was successfully conducted. The heat affected zone of the welding, was about 21 mm width. It was quite adequate to make small specimens, such as SS-3 type sheet tensile specimen.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  4. Microstructure and hardness of HIP-bonded regions in F82H blanket structures

    Science.gov (United States)

    Furuya, K.; Wakai, E.; Ando, M.; Sawai, T.; Nakamura, K.; Takeuchi, H.; Iwabuchi, A.

    2002-12-01

    Metallurgical examinations and hardness measurements were performed at hot isostatic pressing (HIP)-bonded regions in blanket structures made from F82H alloy in order to investigate the HIP-bondability and the influence on the microstructure due to the HIP and heat treatments which would correspond to the fabrication of an actual blanket. The metallurgical examination showed that the HIP-bonded interfaces were sufficiently diffusion-bonded without significant defects, i.e. voids and/or exfoliations, although grain coarsening was observed at a part of the HIP interfaces. Hardness was nearly equal in the coarsening region and a region without coarsening, but about a 10 Hv increase was found in a boundary in between the regions with and without coarsening. Microcrystallized grains were observed in a region about ˜6 μm from HIP interfaces, and the hardness increased by about 0.2 GPa in the region.

  5. On the transition toughness of two RA martensitic steels in the irradiation hardening regime: a mechanism-based evaluation

    Science.gov (United States)

    Odette, G. R.; Rathbun, H. J.; Rensman, J. W.; van den Broek, F. P.

    2002-12-01

    An analysis of the transition fracture toughness and constitutive behavior of F82H and Eurofer97 reduced activation martensitic steels are presented in both unirradiated and irradiated conditions. The unirradiated toughness data for F82H show very steep temperature dependence and the Eurofer97 toughness data measured with 5 mm versus 10 mm thick specimens are systematically higher. Both of these observations indicate a loss of constraint. Constraint loss adjustments are applied using a three-dimensional finite element analysis based toughness scaling model. The adjusted F82H results can be represented by a master curve (MC) and the corresponding 5 and 10 mm adjusted data fall in the same scatter band. The 10 mm irradiated specimens, with generally lower toughness levels, suffer minimal constraint loss. The irradiation induced MC T0 shifts (Δ T0) are analyzed in terms of changes in constitutive properties. The Δ T0 are generally consistent with the observed irradiation hardening. However, the effects of irradiation on post-yield strain hardening behavior must be considered to obtain self-consistent hardening-shift relations.

  6. A master curve analysis of F82H using statistical and constraint loss size adjustments of small specimen data

    Science.gov (United States)

    Odette, G. R.; Yamamoto, T.; Kishimoto, H.; Sokolov, M.; Spätig, P.; Yang, W. J.; Rensman, J.-W.; Lucas, G. E.

    2004-08-01

    We assembled a fracture toughness database for the IEA heat of F82H based on a variety of specimen sizes with a nominal ASTM E1921 master curve (MC) reference temperature T0=-119±3 °C. However, the data are not well represented by a MC. T0 decreases systematically with a decreasing deformation limit Mlim starting at ≈200, which is much higher than the E1921 censoring limit of 30, indicating large constraint loss in small specimens. The small scale yielding T0 at high Mlim is ≈98±5 °C. While, the scatter was somewhat larger than predicted, after model-based adjustments for the effects of constraint loss, the data are in reasonably good agreement with a MC with T0=-98 °C. This supports to use of MC methods to characterize irradiation embrittlement, as long as both constraint loss and statistical size effects are properly accounted for. Finally, we note various issues, including sources of the possible excess scatter, which remain to be fully assessed.

  7. A master curve analysis of F82H using statistical and constraint loss size adjustments of small specimen data

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G.R. E-mail: odette@engineering.ucsb.edu; Yamamoto, T.; Kishimoto, H.; Sokolov, M.; Spaetig, P.; Yang, W.J.; Rensman, J.-W.; Lucas, G.E

    2004-08-01

    We assembled a fracture toughness database for the IEA heat of F82H based on a variety of specimen sizes with a nominal ASTM E1921 master curve (MC) reference temperature T{sub 0}=-119{+-}3 deg. C. However, the data are not well represented by a MC. T{sub 0} decreases systematically with a decreasing deformation limit M{sub lim} starting at {approx}200, which is much higher than the E1921 censoring limit of 30, indicating large constraint loss in small specimens. The small scale yielding T{sub 0} at high M{sub lim} is {approx}98{+-}5 deg. C. While, the scatter was somewhat larger than predicted, after model-based adjustments for the effects of constraint loss, the data are in reasonably good agreement with a MC with T{sub 0}=-98 deg. C. This supports to use of MC methods to characterize irradiation embrittlement, as long as both constraint loss and statistical size effects are properly accounted for. Finally, we note various issues, including sources of the possible excess scatter, which remain to be fully assessed.

  8. Microstructural analysis of ferritic-martensitic steels irradiated at low temperature in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, N.; Robertson, J.P.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Wakai, E. [Japan Atomic Energy Research Inst. (Japan)

    1998-09-01

    Disk specimens of ferritic-martensitic steel, HT9 and F82H, irradiated to damage levels of {approximately}3 dpa at irradiation temperatures of either {approximately}90 C or {approximately}250 C have been investigated by using transmission electron microscopy. Before irradiation, tempered HT9 contained only M{sub 23}C{sub 6} carbide. Irradiation at 90 C and 250 C induced a dislocation loop density of 1 {times} 10{sup 22} m{sup {minus}3} and 8 {times} 10{sup 21} m{sup {minus}3}, respectively. in the HT9 irradiated at 250 C, a radiation-induced phase, tentatively identified as {alpha}{prime}, was observed with a number density of less than 1 {times} 10{sup 20} m{sup {minus}3}. On the other hand, the tempered F82H contained M{sub 23}C{sub 6} and a few MC carbides; irradiation at 250 C to 3 dpa caused minor changes in these precipitates and induced a dislocation loop density of 2 {times} 10{sup 22} m{sup {minus}3}. Difference in the radiation-induced phase and the loop microstructure may be related to differences in the post-yield deformation behavior of the two steels.

  9. Microstructure and microhardness of CLAM steel irradiated up to 20.8 dpa in STIP-V

    Science.gov (United States)

    Peng, Lei; Ge, Hongen; Dai, Yong; Huang, Qunying; Ye, Minyou

    2016-01-01

    Specimens of China low activation martensitic (CLAM) steel were irradiated in the fifth experiment of SINQ target irradiation program (STIP-V) up to 20.8 dpa/1564 appm He. Microhardness measurements and transmission electron microscope (TEM) observations have been performed to investigate irradiation induced hardening effects. The results of CLAM steel specimens show similar trend in microhardness and microstructure changes with irradiation dose, compared to F82H/Optimax-A steels irradiated in STIP-I/II. Defects and helium bubbles were observed in all specimens, even at a very low dose of 5.4 dpa. For defects and bubbles, the mean size and number density increased with increasing irradiation dose to 13 dpa, and then the mean size increased and number density decreased with the increasing irradiation dose to 20.8 dpa.

  10. Final report for the year 2001 on experimental and theoretical investigations of irradiation effects on physical and mechanical properties of iron and RAFM steels

    DEFF Research Database (Denmark)

    Singh, Bachu Narain

    2003-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties have been studied both experimentally and theoretically. Specimens of pure iron and RAFM (reduced activation ferritic-martensic) steels were irradiated todifferent dose levels and at different irradiation...... hardening and formation of “cleared channels” were studied using different computational techniques. Experiments have shown that nano-voids areformed both in pure iron and F82H steel already at 50°C. In pure iron, the formation of nano-voids is detected already at a dose level of ~10-3 dpa. Also in iron...

  11. Technical issues of reduced activation ferritic/martensitic steels for fabrication of ITER test blanket modules

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)], E-mail: tanigawa.hiroyasu@jaea.go.jp; Hirose, T.; Shiba, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kasada, R. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Wakai, E. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Serizawa, H.; Kawahito, Y. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Kimura, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Kohno, Y. [Department of Materials Science and Engineering, Muroran Institute of Technology, Muroran, Hokkaido 050-8585 (Japan); Kohyama, A. [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Katayama, S. [Joining and Welding Research Institute, Osaka University, Ibaraki, Osaka 567-0047 (Japan); Mori, H.; Nishimoto, K. [Division of Materials and Manufacturing Science, Osaka University, Ibaraki, Osaka 565-0871 (Japan); Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6132 (United States)

    2008-12-15

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems. The RAFM F82H was developed in Japan with emphasis on high-temperature properties and weldability. Extensive irradiation studies have conducted on F82H, and it has the most extensive available database of irradiated and unirradiated properties of all RAFMs. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of an ITER test blanket module (TBM) suggested from the recent research achievements in Japan. This work clarified that the primary issues with F82H involve welding techniques and the mechanical properties of weld joints. This is the result of the distinctive nature of the joint caused by the phase transformation that occurs in the weld joint during cooling, and its impact on the design of a TBM will be discussed.

  12. Joining technologies of reduced activation ferritic/martensitic steel for blanket fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [JAERI, Naka Fusion Establishment, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)]. E-mail: hiroset@fusion.naka.jaeri.go.jp; Shiba, K. [JAERI, Naka Fusion Establishment, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan); Ando, M. [JAERI, Naka Fusion Establishment, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan); Enoeda, M. [JAERI, Naka Fusion Establishment, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan); Akiba, M. [JAERI, Naka Fusion Establishment, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

    2006-02-15

    Reduced activation ferritic/martensitic steel, like F82H has been developed as a structural material for in vessel components because of its superior resistance to irradiation damage. As a blanket fabrication process, hot isostatic pressing (HIP) bonding has the great merit of near-net-shaping processing. The degassing conditions and surface roughness were investigated as parameters of HIP conditions. Although the surface roughness and degassing conditions had slight effects on tensile properties, the lack of degassing caused significant degradation of impact properties. A dissimilar metal joint between sintered tungsten and F82H was fabricated by a spark plasma sintering (SPS) method. The joint had no defects in spite of the large difference in thermal expansion coefficient between tungsten and F82H. It is considered that formation of a compliant layer of the ferritic phase can lead to successful bonding for the tungsten and F82H joint even without an artificial interlayer.

  13. Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H. E-mail: tanigawa@popsvr.tokai.jaeri.go.jp; Hashimoto, N.; Sakasegawa, H.; Klueh, R.L.; Sokolov, M.A.; Shiba, K.; Jitsukawa, S.; Kohyama, A

    2004-08-01

    The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile-brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr-2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr-2WVTa)

  14. Metallography studies and hardness measurements on ferritic/martensitic steels irradiated in STIP

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); China Institute of Atomic Energy, Beijing 102413 (China); Long, B. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Dai, Y. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland)], E-mail: yong.dai@psi.ch

    2008-06-30

    In this work metallography investigations and microhardness measurements have been performed on 15 ferritic/martensitic (FM) steels and 6 weld metals irradiated in the SINQ Target Irradiation Program (STIP). The results demonstrate that all the steels have quite similar martensite lath structures. However, the sizes of the prior austenite grain (PAG) of these steels are quite different and vary from 10 to 86 {mu}m. The microstructure in the fusion zones (FZ) of electron-beam welds (EBWs) of 5 steels (T91, EM10, MANET-II, F82H and Optifer-IX) is similar in respect to the martensite lath structure and PAG size. The FZ of the inert-gas-tungsten weld (TIGW) of the T91 steel shows a duplex structure of large ferrite gains and martensite laths. The microhardness measurements indicate that the normalized and tempered FM steels have rather close hardness values. The unusual high hardness values of the EBW and TIGW of the T91 steel were detected, which suggests that these materials are without proper tempering or post-welding heat treatment.

  15. Microstructure and mechanical property in heat affected zone (HAZ in F82H jointed with SUS316L by fiber laser welding

    Directory of Open Access Journals (Sweden)

    S. Kano

    2016-12-01

    Full Text Available This study investigates the microstructure and mechanical property in heat affected zone (HAZ between F82H and SUS316L jointed by 4 kW fiber laser welding at different parameters such as laser scan rate and beam position. OM/FE-SEM observation, EPMA analysis and nano-indentation hardness test were utilized to characterize the microstructure and evaluate the mechanical property. Results show that the HAZ width is dependent on the welding condition. The precipitation of M23C6 particle in HAZ is found to be closely related to the distance from WM/HAZ interface. Decrease in Cr and C concentration in M23C6 depended on the welding condition; the decrease was relatively milder in the case of shifting the beam position to SUS side. Furthermore, the rapid increment in nano-indentation hardness, i.e. ≈2500 MPa, at HAZ/F82H interface was observed regardless of welding parameters. The temperatures at HAZ/F82H interface were estimated from Cr and C concentration change of M23C6 by EPMA. It was revealed that the temperature of HAZ/F82H interface increased with increasing HAZ width, and that the presence of over-tempered HAZ (THAZ region is confirmed only in the specimens welded right on the F82H/SUS interface (no-shift at the laser scan rate of 3 m/min.

  16. Development of an extensive database of mechanical properties for Reduced Activation Ferritic/Martensitic Steels

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, H.; Shiba, K.; Ando, M.; Wakai, E.; Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Kasada, R.; Kimura, A.; Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan); Kohno, Y. [Muroran Institute of Technology, Muroran, Hokkaido (Japan); Klueh, R.L. [0ak Ridge Noational Laboratory, TN (United States); Sokolov, M.; Stoller, R.; Zinklek, S. [0ak Ridge Noational Laboratory, Materials Science and Technology Div., TN (United States); Yamamoto, T.; Odette, G. [UCSB, Dept. of Chemical Engineering UCSB, Santa-Barbara (United States); Kurtz, R.J. [Pacifie Northwest National Laboratory, Richland WA (United States)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H (8Cr-2W-0.2V-0.04Ta-0.1C) and JLF-1 (9Cr-2W-0.2V-0.08Ta-0.1C) are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldablility, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The Japan/US collaboration program also has been conducted with the emphasis on heavy irradiation effects of F82H, JLF-1 and ORNL9Cr2WVTa over the past two decades using Fast Flux Testing Facility (FFTF) of PNNL and High Flux Isotope Reactor (HFIR) of ORNL, and the irradiation condition of the irradiation capsules of those reactors were precisely controlled by the well matured capsule designing and instrumentation. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the database status of RAFMs, mainly on F82H, to identify the key issues for the future development of database. Tensile, fracture toughness, creep and fatigue properties and microstructural studies before and after irradiation are summarized. (authors)

  17. Microstructure evolution of selected ferritic-martensitic steels under dual-beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Wanderka, N.; Camus, E.; Wollenberger, H. [Hahn-Meitner-Inst. Berlin GmbH (Germany)

    1997-11-01

    The authors present experimental results on the microstructure evolution of dual-beam irradiated (300 keV heavy ions plus 15 keV helium ions) ferritic-martensitic steels (Manet, DIN 1.4926, F82H mod). The helium bubble morphology as well as microchemistry of the alloys are investigated by means of transmission electron microscopy and field-ion microscopy with atom probe. The alloys were irradiated to fluences up to 50 dpa and implanted with helium up to a concentration of 1 at.% at the temperatures of 723 K and 773 K. The damage and implantation rates varied from 2.5 {center_dot} 10{sup {minus}3} dpa/s to 2.5 {center_dot} 10{sup {minus}2} dpa/s and from 0.5 appm/s to 5 appm/s, respectively. Size and number density of helium bubbles is found to be rate dependent. Smaller implantation rates produce larger helium bubbles and smaller bubble number densities. Regions of local enrichment of alloy elements, typically 5 nm in size, containing chromium (up to 40 at.%), silicon, and nickel are detected. Number densities of helium bubbles and of regions of chromium enrichments are comparable and lie between 10{sup 23}/m{sup 3} and 10{sup 24}/m{sup 3}. Possible extrapolation of the present ion irradiations to spallation source and fusion reactor conditions is shortly addressed.

  18. Irradiation embrittlement of neutron-irradiated ferritic steel

    Science.gov (United States)

    Kayano, H.; Narui, M.; Ohta, S.; Morozumi, S.

    1985-08-01

    In this study three kinds of Fe-Cr ferritic steels were examined by the instrumented Charpy test and tensile test before and after JMTR irradiation ( 2.2×10 23 f.n./m 2). In the unirradiated samples, 100%-martensite 5Cr-2Mo steel showed the highest adsorbed energy and the highest toughness at low temperatures, follewed by the 9Cr-2Mo steel, and the 20%-martensite 5Cr-2Mo steel showed the third highest toughness. In the irradiated samples, however, thoughness was low as a whole, especially in 20%-martensite 5Cr-2Mo steel. It was clarified that 100%-martensite 5Cr-2Mo steel had the lowest Ductile-to-Brittle Transition Temperature (DBTT) and the highest fracture toughness, and that its DBTT and fracture toughness changed a little upon irradiation, showing excellent irradiation characteristics. The general equations were considered for correlation among strength, ductillity, DBTT and fracture toughness ( J value)

  19. Recent progress in US-Japan collaborative research on ferritic steels R and D

    Energy Technology Data Exchange (ETDEWEB)

    Kimura, Akihiko [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)]. E-mail: Kimura@iae.kyoto-u.ac.jp; Kasada, Ryuta [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kohyama, Akira [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Research Institute, Tokai, Naka-gun, Ibaraki (Japan); Hirose, Takanori [Japan Atomic Energy Research Institute, Tokai, Naka-gun, Ibaraki (Japan); Shiba, Kiyoyuki [Japan Atomic Energy Research Institute, Tokai, Naka-gun, Ibaraki (Japan); Jitsukawa, Shiro [Japan Atomic Energy Research Institute, Tokai, Naka-gun, Ibaraki (Japan); Ohtsuka, Satoshi [Japan Nuclear Cycle Development Institute, Oarai, Higashiibaraki-gun, Ibaraki (Japan); Ukai, Shigeharu [Japan Nuclear Cycle Development Institute, Oarai, Higashiibaraki-gun, Ibaraki (Japan); Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Klueh, R.L. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Yamamoto, Takuya [University of California at Santa Barbara, Santa Barbara, CA (United States); Odette, G.R. [University of California at Santa Barbara, Santa Barbara, CA (United States)

    2007-08-01

    The mechanisms of irradiation embrittlement of two Japanese RAFSs were different from each other. The larger DBTT shift observed in F82H is interpreted by means of both hardening effects and a reduction of cleavage fracture stress by M{sub 23}C{sub 6} carbides precipitation along lath block and packet boundaries, while that of JLF-1 is due to only the hardening effect. Dimensional change measurement during in-pile creep tests revealed the creep strain of F82H was limited at 300 deg C. Performance of the weld bond under neutron irradiation will be critical to determine the life time of blanket structural components. Application of the ODS steels, which are resistant to corrosion in supercritical pressurized water, to the water-cooled blanket is essential to increase thermal efficiency of the blanket systems beyond DEMO. The coupling of RAFS and ODS steel could be effective to realize a highly efficient fusion blanket.

  20. Mechanism of instability of carbides in Fe–TaC alloy under high energy electron irradiation at 673 K

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Hiroaki, E-mail: abe.hiroaki@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Ishizaki, Takahiro [Graduate School of Engineering, Tohoku University, 6-6 Aramaki Aza Aoba, Aoba, Sendai, Miyagi 980-8579 (Japan); Kano, Sho [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Li, Feng [Graduate School of Engineering, Tohoku University, 6-6 Aramaki Aza Aoba, Aoba, Sendai, Miyagi 980-8579 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, 2-1-1 Katahira, Aoba, Sendai, Miyagi 980-8577 (Japan); Tanigawa, Hiroyasu; Hamaguchi, Dai [Japan Atomic Energy Agency, Rokkasho 039-3212 (Japan); Nagase, Takeshi; Yasuda, Hidehiro [Research Center for Ultra-High Voltage Electron Microscopy, Osaka University, 7-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan)

    2014-12-15

    Highlights: • Fe–TaC alloy was fabricated as a model alloy for F82H steel. • Instability of TaC in Fe was observed under high energy electron irradiation at 673 K. • The rate of shrinkage depended on energy, flux, degree of beam focus. • Displacement of Ta in TaC, or radiation-enhanced diffusion of Ta are the mechanism of instability. - Abstract: Reduced activation ferritic/martensitic steels (RAFMs), such as F82H steel, are designed to enhance the high-temperature strength by formation of MX-type nanometer-scale precipitates, mainly TaC. However, their instability under irradiation was recently reported. The purpose of this work, therefore, is to clarify the mechanism employing simultaneous observations under electron irradiation at elevated temperature in a high voltage electron microscope. In this work, Fe-0.2 wt.% TaC was fabricated as a model alloy of F82H steel. The instability of the precipitates was observed under electron irradiation at 1 MeV or above. The remarkable shrinkage and disappearance were clearly observed under irradiation with 1.5 MeV and above. On the contrary, the precipitates were mostly stable below 0.75 MeV. Two kinds of mechanism of the irradiation-induced instability were deduced from the electron-energy dependence. One is the dissolution and diffusion of tantalum from precipitates in ferrite matrix. The other is the displacements of tantalum in precipitates that introduce dissolution of Ta into matrix.

  1. Fracture toughness and Charpy impact properties of several RAFMS before and after irradiation in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, M.A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6151 (United States)]. E-mail: sokolovm@ornl.gov; Tanigawa, H. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Odette, G.R. [University of California-Santa Barbara, Santa Barbara, CA 93106-5080 (United States); Shiba, K. [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Klueh, R.L. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6151 (United States)

    2007-08-01

    As part of the development of candidate reduced-activation ferritic steels for fusion applications, several steels, namely F82H, 9Cr-2WVTa steels and F82H weld metal, are being investigated in the joint DOE-JAEA collaboration program. Within this program, three capsules containing a variety of specimen designs were irradiated at two design temperatures in the ORNL High Flux Isotope Reactor (HFIR). Two capsules, RB-11J and RB-12J, were irradiated in the HFIR removable beryllium positions with europium oxide (Eu{sub 2}O{sub 3}) thermal neutron shields in place. Specimens were irradiated up to 5 dpa. Capsule JP25 was irradiated in the HFIR target position to 20 dpa. The design temperatures were 300 {sup o}C and 500 {sup o}C. Precracked third-sized V-notch Charpy (3.3 x 3.3 x 25.4 mm) and 0.18 T DC(T) specimens were tested to determine transition and ductile shelf fracture toughness before and after irradiation. The master curve methodology was applied to evaluate the fracture toughness transition temperature, T {sub 0}. Irradiation induced shifts of T {sub 0} and reductions of J {sub Q} were compared with Charpy V-notch impact properties. Fracture toughness and Charpy shifts were also compared to hardening results.

  2. Mechanical properties of irradiated 9Cr-2WVTa steel

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States); Rieth, M. [Forschungszentrum Karlsruhe (Germany). Inst. fuer Materialforschung II

    1998-09-01

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

  3. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Summary report

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 250{degrees}C. These specimens have been tested over a temperature range from 20 to 250{degrees}C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenic stainless steels, but the toughness remains quite high. The toughness decreases as the temperature increases. Irradiation at 250{degrees}C is more damaging that at 90{degrees}C, causing larger decreases in the fracture toughness. The ferritic-martensitic steels HT-9 and F82H show significantly greater reductions in fracture toughness that the austenitic stainless steels.

  4. 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.

  5. Mechanism of instability of carbides in Fe-TaC alloy under high energy electron irradiation at 673 K

    Science.gov (United States)

    Abe, Hiroaki; Ishizaki, Takahiro; Kano, Sho; Li, Feng; Satoh, Yuhki; Tanigawa, Hiroyasu; Hamaguchi, Dai; Nagase, Takeshi; Yasuda, Hidehiro

    2014-12-01

    Reduced activation ferritic/martensitic steels (RAFMs), such as F82H steel, are designed to enhance the high-temperature strength by formation of MX-type nanometer-scale precipitates, mainly TaC. However, their instability under irradiation was recently reported. The purpose of this work, therefore, is to clarify the mechanism employing simultaneous observations under electron irradiation at elevated temperature in a high voltage electron microscope. In this work, Fe-0.2 wt.% TaC was fabricated as a model alloy of F82H steel. The instability of the precipitates was observed under electron irradiation at 1 MeV or above. The remarkable shrinkage and disappearance were clearly observed under irradiation with 1.5 MeV and above. On the contrary, the precipitates were mostly stable below 0.75 MeV. Two kinds of mechanism of the irradiation-induced instability were deduced from the electron-energy dependence. One is the dissolution and diffusion of tantalum from precipitates in ferrite matrix. The other is the displacements of tantalum in precipitates that introduce dissolution of Ta into matrix.

  6. An Overview of Irradiation Creep of Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Seog [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper reviewed systematically a state-of-art of irradiation creep for stainless steels to provide a background information for performing irradiation creep tests and establishing the creep model for advanced domestic steels effectively. An irradiation creep model of SFR core materials is necessary to apply to the fuel cladding and assembly materials of domestic SFR reactor system. The document of in-reactor irradiation creep has been obtained by investing a long time and large-scale cost using limited experimental research reactors. This paper will provide the knowledge to understand the irradiation creep and to obtain the background information of advanced domestic steels, so that it hopes to practically apply for timely producing the documents of irradiation creep of advanced domestic steels necessary for the national SFR program.

  7. Irradiation programme HFR phase IIb - SPICE. Impact testing on up to 16.3 dpa irradiated RAFM steels. Final report for task TW2-TTMS 001b-D05

    Energy Technology Data Exchange (ETDEWEB)

    Gaganidze, E.; Dafferner, B.; Ries, H.; Rolli, R.; Schneider, H.C.; Aktaa, J.

    2008-04-15

    The objective of this work is to study the effects of neutron irradiation on the embrittlement behavior of different reduced activation ferritic/martensitic (RAFM) steels. The irradiation was carried out in the Petten High Flux Reactor (HFR) in the framework of the HFR Phase IIb (SPICE) irradiation project at a nominal dose of 16.3 dpa and at different irradiation temperatures (250-450 C). The impact properties are investigated by instrumented Charpy-V tests with subsize specimens (KLST-type). The emphasis is put on the investigation of irradiation induced embrittlement and hardening of the European RAFM reference steel for the first wall of a DEMO fusion reactor, EURO- FER97 under different heat treatment conditions. The mechanical properties of EUROFER97 are compared with the results on international reference steels (F82H-mod, OPTIFER-Ia, GA3X and MANET-I) included in the SPICE project. EUROFER97 irradiated up to 16.3 dpa between 250 and 450 C showed irradiation resistance that is comparable to those of best RAFM steels. Large low temperature (T{sub irr} {<=} 300 C) embrittlement is seen for all investigated RAFM steels. Heat treatment of EUROFER97 at higher austenitizing temperature led to the reduction of embrittlement at low temperatures (T{sub irr} {<=} 350 C). At T{sub irr} {>=} 350 C the DBTTs of the steels remain below -20 C and, hence, are well below the application temperature. Analysis of hardening vs. embrittlement behaviour indicated hardening dominated embrittlement at T{sub irr} {<=} 350 C with 0.17 {<=} C{sub 100} {<=} 0.53 C/MPa. Oxygen dispersion hardened ODS EUROFER with 0.5 wt.% Y{sub 2}O{sub 3} has been irradiated at selected irradiation temperatures. ODS EUROFER showed not satisfying impact properties already in the unirradiated condition characterized by low USE = 2.54 J and large DBTT = 135 C. Furthermore, the increase of USE for irradiation temperatures below T{sub irr} {<=} 350 C indicates not optimized fabrication process. At T{sub irr

  8. Influence of helium on impact properties of reduced-activation ferritic/martensitic Cr-steels

    Science.gov (United States)

    Lindau, R.; Möslang, A.; Preininger, D.; Rieth, M.; Röhrig, H. D.

    Instrumented Charpy impact tests of the reduced activation type 8Cr2WVTa steel F82H have been performed after homogeneous implantation of 300 appm helium at 250°C. The results are compared with investigations on mixed spectrum neutron irradiated (HFR Petten) specimens. After neutron irradiation at 250°C to the same low damage dose of 0.2 dpa, the ductile-brittle transition temperature shift (ΔDBTT) amounts to 18°C, whereas a much higher ΔDBTT of 42°C has been measured after helium implantation. These results are compared with other neutron irradiated ferritic/martensitic steels having different boron levels and thus different helium contents. A model is proposed which describes the dynamic brittle fracture of martensitic/ferritic steels by a stress-induced propagation of micro-cracks, taking into account radiation induced hardening as well as helium bubble formation.

  9. Microstructural change on electron irradiated oxide dispersion strengthened ferritic steels

    Science.gov (United States)

    Kinoshita, H.; Akasaka, N.; Takahashi, H.; Shibahara, I.; Onose, S.

    1992-09-01

    Oxide dispersion strengthened (ODS) ferritic steels were irradiated in a high voltage electron microscope (HVEM) to study their response to irradiation. Fe-13Cr with 0.25 wt% Y2O3 as dispersed particles and containing additions of either 0.45% Nb, 0.45% V and 0.67% Zr were irradiated at 673 and 723 K up to 15 dpa. The Y2O3 particles in all specimens were stable under these irradiation conditions. During irradiation, two types of dislocations were formed but observable voids were not formed. Furthermore, plate-like and granular-like precipitates formed in both the irradiated and nonirradiated regions.

  10. Irradiation behavior of Ti-stabilized 316L type steel

    Science.gov (United States)

    Rodchenkov, B. S.; Kalinin, G. M.; Strebkov, Yu. S.; Shamardin, V. K.; Prokhorov, V. I.; Bulanova, T. M.

    2009-04-01

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose ˜4 and 10 dpa at 265 ± 15 °C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  11. Irradiation behavior of Ti-stabilized 316L type steel

    Energy Technology Data Exchange (ETDEWEB)

    Rodchenkov, B.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, 101000 Moscow (Russian Federation)], E-mail: rodchen@nikiet.ru; Kalinin, G.M.; Strebkov, Yu.S. [Research and Development Institute of Power Engineering (RDIPE), P.O. Box 788, 101000 Moscow (Russian Federation); Shamardin, V.K.; Prokhorov, V.I.; Bulanova, T.M. [State Scientific Center ' Research Institute of Atomic Reactors' , Dimitrovgrad-10, 433510 Ulyanovsk Region (Russian Federation)

    2009-04-30

    Type 316L austenitic steels are widely used for the in-vessel internal structures of fission reactors (core, core support, etc.) and for experimental irradiation facilities. The modifications of 316L Type steel (316L, 316L(N), US 316, J 316, JPCA, etc.) have been considered as structural material for International Thermonuclear Experimental Reactor (ITER). The results of investigation the irradiation behaviour of Ti-stabilized 316 L type steel (0.04 C-15 Cr-11 Ni-2.5 Mo-0.5 Ti) are presented in this work. The specimens cut out from 316L-Ti steel forging were irradiated in the SM-2 reactor up to a dose {approx}4 and 10 dpa at 265 {+-} 15 deg. C. The tensile properties, fracture toughness and changes in resistance to intergranular stress corrosion cracking (IGSCC) have been investigated after irradiation. The results for Ti-stabilized 316L steel were compared with those for 316L(N)-IG steel irradiated at the same condition.

  12. Decomposition of the MANET steel under dual-beam irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Wanderka, N. [Hahn-Meitner-Institut Berlin (Germany); Camus, E. [Hahn-Meitner-Institut Berlin (Germany); Naundorf, V. [Hahn-Meitner-Institut Berlin (Germany); Keilonat, C. [Hahn-Meitner-Institut Berlin (Germany); Welzel, S. [Hahn-Meitner-Institut Berlin (Germany); Wollenberger, H. [Hahn-Meitner-Institut Berlin (Germany)

    1996-02-01

    Decomposition of the MANET steel was observed by means of atom probing after 300 keV Fe{sup +} ion irradiation to 50 dpa and simultaneous implantation of 15 keV He{sup +} ions at a rate of 200 appm/dpa. At irradiation temperatures of 673 and 698 K weak periodical variation of the chromium concentration was observed. At irradiation temperatures of 723 and 773 K clusters with chromium concentration of up to 25 at% were detected. (orig.).

  13. Development of PIE techniques for irradiated LWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-09-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  14. Corrosion and stress corrosion cracking of ferritic/martensitic steel in super critical pressurized water

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [Naka Fusion Research Institute, JAEA, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)]. E-mail: hirose.takanori@jaea.go.jp; Shiba, K. [Naka Fusion Research Institute, JAEA, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan); Enoeda, M. [Naka Fusion Research Institute, JAEA, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan); Akiba, M. [Naka Fusion Research Institute, JAEA, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

    2007-08-01

    A water-cooled solid breeder (WCSB) blanket cooled by high temperature SCPW (super critical pressurized water) is a practical option of DEMO reactor. Therefore, it is necessary to check the compatibility of the steel with SCPW. In this work, reduced activation ferritic/martensitic steel, F82H has been tested through slow strain rate tests (SSRT) in 23.5 MPa SCPW. And weight change behavior was measured up to 1000 h. F82H did not demonstrated stress corrosion cracking and its weight simply increased with surface oxidation. The weight change of F82H was almost same as commercial 9%-Cr steels. According to a cross-sectional analysis and weight change behavior, corrosion rate of F82H in the 823 K SCPW is estimated to be 0.04 mm/yr.

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

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Kiyohiro, E-mail: kiyohiro.yabuuchi@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Kuribayashi, Yutaka [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Nogami, Shuhei, E-mail: shuhei.nogami@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hasegawa, Akira, E-mail: akira.hasegawa@qse.tohoku.ac.jp [Graduate School of Engineering, Tohoku University, 6-6-01-2 Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan)

    2014-03-15

    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{sup +} 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{sub 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{sub 0}. Thus, the nanoindentation technique demonstrated in this study is valuable for measuring irradiation hardening in ion-irradiated materials.

  16. 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.

  17. Final report on neutron irradiation at low temperature to investigate plastic instability and at high temperature to study caviation

    DEFF Research Database (Denmark)

    Singh, B.N; Eldrup, Morten Mostgaard; Golubov, D.J.

    2005-01-01

    Effects of neutron irradiation on defect accumulation and physical and mechanical properties of pure iron and F82H and EUROFER 97 ferritic-martensitic steels have been investigated. Tensile specimens were neutron irradiated to a dose level of 0,23 dpa at333 and 573 K. Electrical resistivity...... and tensile properties were measured both in the unirradiated and irradiated condition. Some additional specimens of pure iron were irradiated at 333 K to doses of 10-3, 10-2 and 10-1 dpa and tensile tested at 333 K.To investigate the effect of helium on cavity nucleation and growth, specimens of pure iron...... and EUROFER 97 were implanted with different amounts of helium at 323 K and subsequently neutron irradiated to doses of 10-3, 10-2 and 10-1 dpa at 323 K. Defectmicrostructures were investigated using positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM). Numerical calculations...

  18. The studies of irradiation hardening of stainless steel reactor internals under proton and xenon irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Chaoliang; Zhang, Lu; Qian, Wangjie; Mei, Jinna; Liu, Xiang Bing [Suzhou Nuclear Power Research Institute, Suzuhou (China)

    2016-06-15

    Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

  19. Defect structure of irradiated PH13-8Mo steel

    Science.gov (United States)

    Van Renterghem, W.; Al Mazouzi, A.; Van den Berghe, S.

    2007-02-01

    PH13-8Mo bolts, which are considered for use in the ITER reactor, were irradiated up to doses of 0.5, 1 and 2 dpa. The microstructure was investigated with transmission electron microscopy and its evolution is discussed with reference to the mechanical properties. PH13-8Mo is a precipitation hardened martensitic steel, but a large amount of austenite has been observed as well. The precipitation hardening results from the formation of small coherent NiAl precipitates in the martensite phase. Their size, size distribution and density are found to be unaffected by neutron irradiation. The dislocations in the martensite phase are mainly a/2 type screw dislocations, whereas in the austenite phase mainly a/2 type screw dislocations are present. The line dislocation structure did not change during irradiation, but small irradiation induced defects were observed. Using the Orowan model, it is argued that the latter are responsible for the irradiation hardening.

  20. Sensitivity of ultrasonic nonlinearity to irradiated, annealed, and re-irradiated microstructure changes in RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Matlack, K.H., E-mail: katie.matlack@gatech.edu [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Kim, J.-Y. [School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Wall, J.J. [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Electric Power Research Institute, Charlotte, NC 28262 (United States); Qu, J. [Department of Civil and Environmental Engineering, Northwestern University, Evanston, IL 60208 (United States); Jacobs, L.J. [G.W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); School of Civil and Environmental Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States); Sokolov, M.A. [Metals and Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The planned life extension of nuclear reactors throughout the US and abroad will cause reactor vessel and internals materials to be exposed to more neutron irradiation than was originally intended. A nondestructive evaluation (NDE) method to monitor radiation damage would enable safe and cost-effective continued operation of nuclear reactors. Radiation damage in reactor pressure vessel (RPV) steels causes microstructural changes that leave the material in an embrittled state. Nonlinear ultrasound is an NDE technique quantified by the measurable acoustic nonlinearity parameter, which is sensitive to microstructural changes in metallic materials such as dislocations, precipitates and their combinations. Recent research has demonstrated the sensitivity of the acoustic nonlinearity parameter to increasing neutron fluence in representative RPV steels. The current work considers nonlinear ultrasonic experiments conducted on similar RPV steel samples that had a combination of irradiation, annealing, re-irradiation, and/or re-annealing to a total neutron fluence of 0.5–5 × 10{sup 19} n/cm{sup 2} (E > 1 MeV) at an irradiation temperature of 290 °C. The acoustic nonlinearity parameter generally increased with increasing neutron fluence, and consistently decreased from the irradiated to the annealed state over different levels of neutron fluence. Results of the measured acoustic nonlinearity parameter are compared with those from previous measurements on other RPV steel samples. This comprehensive set of results illustrates the dependence of the measured acoustic nonlinearity parameter on neutron fluence, material composition, irradiation temperature and annealing.

  1. Sensitivity of ultrasonic nonlinearity to irradiated, annealed, and re-irradiated microstructure changes in RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Matlack, Katie [Georgia Institute of Technology, Atlanta; Kim, J-Y. [Georgia Institute of Technology, Atlanta; Wall, J.J. [Electric Power Research Institute (EPRI); Jacobs, L.J. [Georgia Institute of Technology, Atlanta; Sokolov, Mikhail A [ORNL

    2014-05-01

    The planned life extension of nuclear reactors throughout the US and abroad will cause reactor vessel and internals materials to be exposed to more neutron irradiation than was originally intended. A nondestructive evaluation (NDE) method to monitor radiation damage would enable safe and cost-effective continued operation of nuclear reactors. Radiation damage in reactor pressure vessel (RPV) steels causes microstructural changes that leave the material in an embrittled state. Nonlinear ultrasound is an NDE technique quantified by the measurable acoustic nonlinearity parameter, which is sensitive to microstructural changes in metallic materials such as dislocations, precipitates and their combinations. Recent research has demonstrated the sensitivity of the acoustic nonlinearity parameter to increasing neutron fluence in representative RPV steels. The current work considers nonlinear ultrasonic experiments conducted on similar RPV steel samples that had a combination of irradiation, annealing, re-irradiation, and/or re-annealing to a total neutron fluence of 0.5 5 1019 n/cm2 (E > 1 MeV) at an irradiation temperature of 290 C. The acoustic nonlinearity parameter generally increased with increasing neutron fluence, and consistently decreased from the irradiated to the annealed state over different levels of neutron fluence. Results of the measured acoustic nonlinearity parameter are compared with those from previous measurements on other RPV steel samples. This comprehensive set of results illustrates the dependence of the measured acoustic nonlinearity parameter on neutron fluence, material composition, irradiation temperature and annealing.

  2. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  3. Mechanical and physical properties of irradiated type 348 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Beeston, J.M.

    1980-01-01

    A type 348 stainless steel in-pile tube irradiated to a fluence of 3 x 10/sup 22/ n/cm/sup 2/, E > 1 MeV (57 dpa), was destructively examined. The service had resulted in a maximum total creep of 1.8% at the high fluence. The metal temperature ranged between 623 and 652/sup 0/K, hence the thermal creep portion of the total was negligible. Total creep was greater than had been anticipated from creep data for austenitic stainless steels irradiated in other reactors. The objectives of the destructive examination were to determine the service-induced changes of mechanical and physical properties, and to assess the possibility of adverse effects of both these changes and the greater total creep on the prospective service life of other tubes.

  4. Irradiation programme MANITU. Results of impact tests with the irradiated materials of the first irradiation phase (0.8 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 0,8 dpa bestrahlten Werkstoffen der ersten Bestrahlungsphase

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Dafferner, B.; Ries, H.; Romer, O.

    1995-09-01

    The irradiation project MANITU was planned and carried out in the frame of the European Longterm Fusion Materials Development Programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of subsize Charpy tests with the unirradiated reference specimens of MANITU a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are partly better compared to the modified commercial 10-11% Cr-NiMoVNb steels. In the present report the results of instrumented impact tests within the first phase of the MANITU programme (irradiation dose 0.8 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed. Among all examined alloys (MANET-I, MANET-II, K-heat, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) the ORNL steel shows the very best embrittlement behaviour after neutron irradiation. (orig.) [Deutsch] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant und durchgefuehrt. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnete sich jedoch nach der Auswertung der Kerbschlagbiegeversuche an den unbestrahlten miniaturisierten Referenzproben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-W(Ge)VTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen teilweise bessere mechanische Eigenschaften auf. Im vorliegenden Bericht werden die Ergebnisse aus den instrumentierten Kerbschlagbiegeversuchen der ersten Phase des MANITU-Programms (Bestrahlungsdosis 0,8 dpa, Bestrahlungstemperaturen 250, 300, 350, 400 und 450 C) analysiert und bewertet. Von den untersuchten Legierungen (MANET-I, MANET-II, Kastencharge, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791

  5. Materials irradiation facilities at the high-power Swiss proton accelerator complex

    Science.gov (United States)

    Wagner, Werner; Dai, Yong; Glasbrenner, Heike; Aebersold, Hans-Ulrich

    2007-04-01

    Within the Swiss proton accelerator complex at the Paul-Scherrer-Institute (PSI), several irradiation facilities are operated for investigation of materials behavior under high-dose irradiation conditions as well as for neutron activation analysis and isotope production. In LiSoR (liquid solid reaction), a liquid metal loop connected to the 72 MeV proton accelerator Injector 1, steel samples are irradiated while being in contact with flowing lead-bismuth-eutectic (LBE) at elevated temperatures and under tensile stress. In the spallation neutron source SINQ, the STIP program (SINQ Target Irradiation Program) allows materials irradiation under realistic spallation conditions, i.e. in a mixed spectrum of 570 MeV protons and spallation neutrons. Hundreds of samples, mainly austenitic and ferritic-martensitic steels such as 316L, T91 or F82H, were irradiated to doses up to 20 dpa as part of STIP. These also included steel samples in contact with liquid Hg and liquid LBE. MEGAPIE (MEGAwatt PIlot Experiment), a liquid metal target employing LBE, operated in SINQ during the second half of 2006, can be taken as a materials irradiation facility on its own. Adjacent to the target position, SINQ houses a neutron irradiation rabbit system serving activation analysis and isotope production.

  6. a Study of Stress Relaxation Rate in Un-Irradiated and Neutron-Irradiated Stainless Steel

    Science.gov (United States)

    Ghauri, I. M.; Afzal, Naveed; Zyrek, N. A.

    Stress relaxation rate in un-irradiated and neutron-irradiated 303 stainless steel was investigated at room temperature. The specimens were exposed to 100 mC, Ra-Be neutron source of continuous energy 2-12 MeV for a period ranging from 4 to 16 days. The tensile deformation of the specimens was carried out using a Universal Testing Machine at 300 K. During the deformation, straining was frequently interrupted by arresting the cross head to observe stress relaxation at fixed load. Stress relaxation rate, s, was found to be stress dependent i.e. it increased with increasing stress levels σ0 both in un-irradiated and irradiated specimens, however the rate was lower in irradiated specimens than those of un-irradiated ones. A further decrease in s was observed with increase in exposure time. The experiential decrease in the relaxation rate in irradiated specimens is ascribed to strong interaction of glide dislocations with radiation induced defects. The activation energy for the movement of dislocations was found to be higher in irradiated specimens as compared with the un-irradiated ones.

  7. Oxide dispersion strengthened steel irradiation with helium ions

    Energy Technology Data Exchange (ETDEWEB)

    Pouchon, M.A. [Laboratory for Materials Behaviour, Paul Scherrer Institute, OHLA/131, 5232 Villigen PSI (Switzerland)]. E-mail: manuel.pouchon@psi.ch; Chen, J. [Laboratory for Materials Behaviour, Paul Scherrer Institute, OHLA/131, 5232 Villigen PSI (Switzerland); Doebeli, M. [Laboratory for Materials Behaviour, Paul Scherrer Institute, OHLA/131, 5232 Villigen PSI (Switzerland); Hoffelner, W. [Laboratory for Materials Behaviour, Paul Scherrer Institute, OHLA/131, 5232 Villigen PSI (Switzerland)

    2006-06-30

    Oxide dispersion strengthened (ODS) ferritic steels are investigated as possible structural material for the future generation of high temperature gas cooled nuclear reactors. ODS-steels are considered to replace other high temperature materials for tubing or structural parts. The oxide particles serve for interfacial pinning of moving dislocations. Therefore, the creep resistance is improved. In case of the usage of these materials in reactors, the behavior under irradiation must be further clarified. In this paper the effects induced by {sup 4}He{sup 2+} implantation into a ferritic ODS steel are investigated. The fluence ranges from 10{sup 16} to 10{sup 17} cm{sup -2} and the energy from 1 to 2 MeV. The induced swelling is investigated for implantations at room temperature and 470 K. It is derived from the irradiation induced surface displacement, which is measured with an atomic force microscope (AFM). With a displacement damage of 0.6 dpa, a volume increase of 0.65% is observed at room temperature and 0.33% at 470 K. A cross-sectional cut is performed by focused ion beam and investigated by transmission electron microcopy (TEM). The defect density observed on the TEM micrographs agrees well with the computational simulation (TRIM) of the damage profile.

  8. Defect structure of irradiated PH13-8Mo steel

    Energy Technology Data Exchange (ETDEWEB)

    Renterghem, W. van [SCK-CEN, Reactor Materials Research, Boeretang 200, 2400 Mol (Belgium)]. E-mail: wvrenter@sckcen.be; Al Mazouzi, A. [SCK-CEN, Reactor Materials Research, Boeretang 200, 2400 Mol (Belgium); Berghe, S. van den [SCK-CEN, Reactor Materials Research, Boeretang 200, 2400 Mol (Belgium)

    2007-02-01

    PH13-8Mo bolts, which are considered for use in the ITER reactor, were irradiated up to doses of 0.5, 1 and 2 dpa. The microstructure was investigated with transmission electron microscopy and its evolution is discussed with reference to the mechanical properties. PH13-8Mo is a precipitation hardened martensitic steel, but a large amount of austenite has been observed as well. The precipitation hardening results from the formation of small coherent NiAl precipitates in the martensite phase. Their size, size distribution and density are found to be unaffected by neutron irradiation. The dislocations in the martensite phase are mainly a/2<1 1 1> type screw dislocations, whereas in the austenite phase mainly a/2<1 1 0> type screw dislocations are present. The line dislocation structure did not change during irradiation, but small irradiation induced defects were observed. Using the Orowan model, it is argued that the latter are responsible for the irradiation hardening.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

  10. Influence of Simulated Outside-Reactor Irradiation on Anticorrosion Property of Austenitic Stainless Steel

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The influence of γ-ray irradiation on the properties of inside-reactor stainless steel structures was studied by simulating the working condition of pressurized water reactor (PWR) first circuit and the outside-reactor γ-ray irradiation. The result shows that the simulated outside-reactor irradiation (irradiation dose 4.4 × 104 Gy) has no influence on anticorrosion properties of solutionized SUS304 austenitic stainless steel, including intergranular corrosion (IC) and stress corrosion cracking (SCC). Anticorrosion properties (IC, SCC) of sensitized SUS304 austenitic stainless steel are reduced by simulated outside-reactor irradiation. The longer the sensitizedtime is, the more obvious the influence is.

  11. Dependence of Radiation Damage in Stainless Steel on Irradiation Dose

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    The accelerator driven radioactive clean nuclear power system (ADS) is a novel innovative idea forthe sustainable development of nuclear power system. The spallation neutron source system is one of thethree key parts of ADS, which provides source neutrons of about 1018 s-1 for the burning-up of fuels.Stainless steel (SS) is used for the beam window and target materials of the spallation neutron sourcesystem. It is irradiated by high-energy and intense protons and/or neutrons during operation. Theaccumulated displacement damage dose could reach a couple of hundred dpa (displacement per atom) per

  12. Nanostructure evolution in ODS steels under ion irradiation

    Directory of Open Access Journals (Sweden)

    S. Rogozhkin

    2016-12-01

    In this work, we carried out atom probe tomography (APT and transmission electron microscopy (TEM studies of three different ODS steels produced by mechanical alloying: ODS Eurofer, 13.5Cr ODS and 13.5Cr-0.3Ti ODS. These materials were investigated after irradiation with Fe (5.6MeV or Ti (4.8MeV ions up to 1015ion/cm2 and part of them up to 3×1015ion/cm2. In all cases, areas for TEM investigation were cut at a depth of ∼ 1.3µm from the irradiated surface corresponding to the peak of the radiation damage dose. It was shown that after irradiation at RT and at 300°С the number density of oxide particles in all the samples grew up. Meanwhile, the fraction of small particles in the size distribution has increased. APT revealed an essential increase in nanoclusters number and a change of their chemical composition at the same depth. The nanostructure was the most stable in 13.5Cr-0.3Ti ODS irradiated at 300°С: the increase of the fraction of small oxides was minimal and no change of nanocluster chemical composition was detected.

  13. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    In order to investigate the role of He in cavity nucleation in neutron irradiated iron and steel, pure iron and Eurofer-97 steel have been He implanted and neutron irradiated in a systematic way at different temperatures, to different He and neutron doses and with different He implantation rates...

  14. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States)]. E-mail: kluehrl@ornl.gov; Hashimoto, N. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States); Sokolov, M.A. [Oak Ridge National Laboratory, P.O. Box 2008, MS 6151, Oak Ridge, Tennessee 37831-6151 (United States); Shiba, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan); Jitsukawa, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan)

    2006-10-15

    Tensile and Charpy specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 deg. C in the High Flux Isotope Reactor (HFIR) up to {approx}12 dpa and at 393 deg. C in the Fast Flux Test Facility (FFTF) to {approx}15 dpa. In HFIR, a mixed-spectrum reactor (n, {alpha}) reactions of thermal neutrons with {sup 58}Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile-brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 deg. C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2-4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. The effect of oxygen on void stability in ion-irradiated steel

    Science.gov (United States)

    Seitzman, Larry E.; Dodd, R. Arthur; Kulcinski, Gerald L.

    1990-07-01

    The effect of oxygen on void stability in an Fe-17Ni-13Cr austenitic ternary alloy has been investigated using 15 MeV nickel-ion irradiation at elevated temperatures and preimplantation of 6 MeV oxygen at room temperature. The nickel irradiation was performed over a temperature range of 550 °C to 650 °C. Utilizing transverse specimen preparation techniques, the irradiated steel was examined by transmission electron microscopy (TEM). As little as 10 appm preimplanted oxygen caused a significant increase in the void number density when the steel was irradiated at 550 °C. A near-surface void-denuded zone occurs in the irradiated steel, while a region depleted of visible voids also occurs in the steel injected with 300 appm oxygen or greater and irradiated at 550 °C.

  17. Irradiation programme MANITU. Results of impact tests with the irradiated materials (0.2 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 0,2 dpa bestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Rieth, M.; Dafferner, B.; Kunisch, W.; Ries, H.; Romer, O.

    1997-04-01

    The irradiation project MANITU was planned and carried out in the frame of the European Longterm Fusion Materials Development Programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of subsize Charpy tests with specimens of MANITU (0.8 dpa) a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are better compared to the modified commercial 10-11% Cr-NiMoVNb steels. In the present report the results of instrumented impact tests of the MANITU programme (irradiation dose 0.2 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed. Among all examined alloys (MANET-I, MANET-II, K-heat, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) the ORNL steel shows the best embrittlement behaviour after neutron irradation. (orig.) [Deutsch] Das Bestrahlungsprojekt MANITU wurde im Rahmen des europaeischen Langzeitprogramms fuer Materialentwicklung fuer die Kernfusion geplant und durchgefuehrt. Das Problem der bestrahlungsinduzierten Versproedung bei den in Frage kommenden martensitischen Werkstoffen ist nach wie vor ungeloest. Eine erste Tendenz zeichnete sich jedoch nach der Auswertung der Kerbschlagbiegeversuche an den bis 0,8 dpa bestrahlten miniaturisierten Proben des MANITU-Programms ab. Die neu entwickelten niedrig aktivierbaren 7-10% Cr-W(Ge)VTa-Legierungen weisen gegenueber den modifizierten kommerziellen 10-11% Cr-NiMoVNb-Staehlen bessere mechanische Eigenschaften auf. Im vorliegenden Bericht werden die Ergebnisse aus den instrumentierten Kerbschlagbiegeversuchen des MANITU-Programms (Bestrahlungsdosis 0,2 dpa, Bestrahlungstemperaturen 250, 300, 350, 400 und 450 C) analysiert und bewertet. Von den untersuchten Legierungen (MANET-I, MANET-II, Kastencharge, OPTIFER-Ia, OPTIFER-II, F82H, 9Cr-2WVTa ORNL 3791) zeigt der ORNL-Stahl das beste Versproedungsverhalten nach

  18. Evaluation of irradiation assisted stress corrosion cracking (IASCC) of type 316 stainless steel irradiated in FBR

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, T. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Jitsukawa, S. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Shiba, K. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Sato, Y. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)); Shibahara, I. (Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)); Nakajima, H. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1993-12-01

    Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOYO was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-induced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425 C to a neutron fluence of 8.3x10[sup 26] n/m[sup 2] (> 0.1 MeV) which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300 C, but was not observed on specimen tested in water at 60 C and in air at 300 C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCC) in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum. (orig.)

  19. Effect of heat treatment and irradiation temperature on impact behavior of irradiated reduced-activation ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1998-03-01

    Charpy tests were conducted on eight normalized-and-tempered reduced-activation ferritic steels irradiated in two different normalized conditions. Irradiation was conducted in the Fast Flux Test Facility at 393 C to {approx}14 dpa on steels with 2.25, 5, 9, and 12% Cr (0.1% C) with varying amounts of W, V, and Ta. The different normalization treatments involved changing the cooling rate after austenitization. The faster cooling rate produced 100% bainite in the 2.25 Cr steels, compared to duplex structures of bainite and polygonal ferrite for the slower cooling rate. For both cooling rates, martensite formed in the 5 and 9% Cr steels, and martensite with {approx}25% {delta}-ferrite formed in the 12% Cr steel. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy. The difference in microstructure in the low-chromium steels due to the different heat treatments had little effect on properties. For the high-chromium martensitic steels, only the 5 Cr steel was affected by heat treatment. When the results at 393 C were compared with previous results at 365 C, all but a 5 Cr and a 9 Cr steel showed the expected decrease in the shift in DBTT with increasing temperature.

  20. Development of benchmark reduced activation ferritic/martensitic steels for fusion energy applications

    Science.gov (United States)

    Tanigawa, H.; Gaganidze, E.; Hirose, T.; Ando, M.; Zinkle, S. J.; Lindau, R.; Diegele, E.

    2017-09-01

    Reduced-activation ferritic/martensitic (RAFM) steel is the benchmark structural material for in-vessel components of fusion reactor. The current status of RAFM developments and evaluations is reviewed based on two leading RAFM steels, F82H and EUROFER-97. The applicability of various joining technologies for fabrication of fusion first wall and blanket structures, such as weld or diffusion bonding, is overviewed as well. The technical challenges and potential risks of utilizing RAFM steels as the structural material of in-vessel components are discussed, and possible mitigation methodology is introduced. The discussion suggests that deuterium-tritium fusion neutron irradiation effects currently need to be treated as an ambiguity factor which could be incorporated within the safety factor. The safety factor will be defined by the engineering design criteria which are not yet developed with regard to irradiation effects and some high temperature process, and the operating time condition of the in-vessel component will be defined by the condition at which those ambiguities due to neutron irradiation become too large to be acceptable, or by the critical condition at which 14 MeV fusion neutron irradiation effects is expected to become different from fission neutron irradiation effects.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-06

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

  2. Technical issues related to the development of reduced-activation ferritic/martensitic steels as structural materials for a fusion blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hiroyasu, E-mail: tanigawa.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Shiba, Kiyoyuki; Sakasegawa, Hideo; Hirose, Takanori; Jitsukawa, Shiro [Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan)

    2011-10-15

    Reduced activation ferritic/martensitic (RAFM) steels are recognized as the primary candidate structural materials for fusion blanket systems. Because of the possibility of creating sound engineering bases, such as a suitable fabrication technology and a materials database, RAFM steels can be used as structural materials for pressure equipment. Further, the development of an irradiation database in addition to design methodologies for fusion-centered applications is critical when evaluating the applicability of RAFM steels as structural materials for fusion-neutron-irradiated pressure equipment. In the International Fusion Energy Research Centre (IFERC) project in the Broader Approach (BA) activities between the EU and Japan, R and D is underway to optimize RAFM steel fabrication and processing technologies, develop a method for estimating fusion-neutron-irradiation effects, and study the deformation behaviors of irradiated structures. The results of these research activities are expected to form the basis for the DEMO power plant design criteria and licensing. The objective of this paper is to review the BA R and D status of RAFM steel development in Japan, especially F82H (Fe-8Cr-2W-V, Ta). The key technical issues relevant to the design and fabrication of the DEMO blanket and the recent achievements in Japan are introduced.

  3. Microstructure and nanoindentation of the CLAM steel with nanocrystalline grains under Xe irradiation

    Science.gov (United States)

    Chang, Yongqin; Zhang, Jing; Li, Xiaolin; Guo, Qiang; Wan, Farong; Long, Yi

    2014-12-01

    This work presents an early look at irradiation effects on China low activation martensitic (CLAM) steel with nanocrystalline grains (NC-CLAM steels) under 500 keV Xe-ion bombardment at room temperature to doses up to 5.3 displacements per atom (dpa). The microstructure in the topmost region of the steel is composed of nanocrystalline grains with an average diameter of 13 nm. As the samples were implanted at low dose, the nanocrystalline grains had martensite lath structure, and many dislocations and high density bubbles were introduced into the NC-CLAM steels. As the irradiation dose up to 5.3 dpa, a tangled dislocation network exists in the lath region, and the size of the bubbles increases. X-ray diffraction results show that the crystal quality decreases after irradiation, although the nanocrystals obviously coarsen. Grain growth under irradiation may be ascribed to the direct impact of the thermal spike on grain boundaries in the NC-CLAM steels. In irradiated samples, a compressive stress exists in the surface layer because of grain growth and irradiation-introduced defects, while the irradiation introduced grain-size coarsening and defects gradients from the surface to matrix result in a tensile stress in the irradiated NC-CLAM steels. Nanoindentation was used to estimate changes in mechanical properties during irradiation, and the results show that the hardness of the NC-CLAM steels increases with increasing irradiation dose, which was ascribed to the competition between the grain boundaries and the irradiation-introduced defects.

  4. Irradiation programme MANITU: results of impact tests with the irradiated materials (2,4 dpa); Bestrahlungsprogramm MANITU. Ergebnisse der Kerbschlagbiegeversuche mit den bis 2,4 dpa bestrahlten Werkstoffen

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, H.C.; Dafferner, B.; Ries, H.; Romer, O.

    2001-05-01

    The irradiation project MANITU was planned and carried out in the frame of the European long-term fusion materials development programme. The problem of the irradiation induced embrittlement of possible martensitic alloy candidates is still unsolved. But after the evaluation of sub-size Charpy tests with the unirradiated reference specimens of MANITU a first tendency was recognizable. The mechanical properties of the newly developed low activation 7-10% Cr-W(Ge)VTa alloys are partly better compared to the modified commercial 10-11% Cr-NiMoVNb steels. After the evaluation of subsize Charpy tests with specimens irradiated up to 0.2 and 0.8 dpa in the first phase of the MANITU programme, better mechanical properties of the 7-10% Cr-W(Ge)VTa alloys were obvious. In the present report the results of instrumented impact tests within the second phase of the MANITU programme (irradiation dose 2.4 dpa, irradiation temperatures 250, 300, 350, 400, and 450 C) are analysed and assessed in comparison to the results of the irradiation up to 0.2 and 0.8 dpa in the first phase of the project. Among the examined alloys (MANET-I/II, K-Heat, OPTIFER-Ia/II, F82H, ORNL 3791) the ORNL steel shows the very best embrittlement behaviour after neutron irradiation. (orig.)

  5. Tensile and charpy impact properties of irradiated reduced-activation ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1996-10-01

    Tensile tests were conducted on eight reduced-activation Cr-W steels after irradiation to 15-17 and 26-29 dpa, and Charpy impact tests were conducted on the steels irradiated to 26-29 dpa. Irradiation was in the Fast Flux Test Facility at 365{degrees}C on steels containing 2.25-12% Cr, varying amounts of W, V, and Ta, and 0.1%C. Previously, tensile specimens were irradiated to 6-8 dpa and Charpy specimens to 6-8, 15-17, and 20-24 dpa. Tensile and Charpy specimens were also thermally aged to 20000 h at 365{degrees}C. Thermal aging had little effect on the tensile behavior or the ductile-brittle transition temperature (DBTT), but several steels showed a slight increase in the upper-shelf energy (USE). After {approx}7 dpa, the strength of the steels increased and then remained relatively unchanged through 26-29 dpa (i.e., the strength saturated with fluence). Post-irradiation Charpy impact tests after 26-29 dpa showed that the loss of impact toughness, as measured by an increase in DBTT and a decrease in the USE, remained relatively unchanged from the values after 20-24 dpa, which had been relatively unchanged from the earlier irradiations. As before, the two 9Cr steels were the most irradiation resistant.

  6. Mechanical properties and microstructure of advanced ferritic-martensitic steels used under high dose neutron irradiation

    Science.gov (United States)

    Shamardin, V. K.; Golovanov, V. N.; Bulanova, T. M.; Povstianko, A. V.; Fedoseev, A. E.; Goncharenko, Yu. D.; Ostrovsky, Z. E.

    Some results of the study of mechanical properties and structure of ferritic-martensitic chromium steels with 13% and 9% chromium, irradiated in the BOR-60 reactor up to different damage doses are presented in this report. Results concerning the behaviour of commercial steels, containing to molybdenum, vanadium and niobium, and developed for the use in fusion reactors, are compared to low-activation steels in which W and Ta replaced Mo and Nb. It is shown that after irradiation to the dose of ˜10 dpa at 400°C 0.1C-9Cr-1W, V, Ta steels are prone to lower embrittlement as deduced from fracture surface observations of tensile specimens. Peculiarities of fine structure and fracture mode, composition and precipitation reactions in steels during irradiation are discussed.

  7. Views of TAGSI on the effects of gamma irradiation on the mechanical properties of irradiated ferritic steel reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Engineering, Metallurgy and Materials, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); English, C.A. [Materials and Chemistry Consultancy, Nexia Solutions, 168 Harwell International Business Centre, Didcot, Oxon OX11 0QJ (United Kingdom); Weaver, D.R. [School of Physics, University of Birmingham, Edgbaston, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Assurance, Walton House, 404 The Quadrant, Birchwood Park, Warrington, Cheshire WA3 6AT (United Kingdom)

    2005-12-01

    The paper reviews and analyses the effects of gamma irradiation dose on the properties of ferritic steels used in reactor pressure vessels (RPVs). It explains factors that affect the embrittlement of a RPV steel induced by combinations of fast neutrons, thermal neutrons, and gamma irradiation. TAGSI were asked to consider the effects of gamma irradiation dose on the properties of steels used in reactor pressure vessels. TAGSI endorsed the use of the MCBEND code to calculate gamma fluxes and energetic gamma ray displacement cross-sections calculated using either Baumann or Alexander methods. TAGSI endorsed the calculation of the materials property changes due to an additional gamma dose using trend curves based on empirical correlation to neutron-induced damage (where k {sub {gamma}}{approx}1{+-}0.25)

  8. Effects of water and irradiation temperatures on IASCC susceptibility of type 316 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi E-mail: ttsukada@popsvr.tokai.jaeri.go.jp; Miwa, Yukio; Jitsukawa, Shiro; Shiba, Kiyoyuki; Ouchi, Asao

    2004-08-01

    Effects of water and irradiation temperatures on irradiation-assisted stress corrosion cracking (IASCC) of type 316 stainless steel were investigated. Type 316 stainless steel was irradiated at 333-673 K to a dose level of 16 dpa. Susceptibility to IASCC was evaluated by slow strain rate testing in oxygenated water in the temperature range of 513-573 K. Irradiation at 603 and 673 K caused IASCC in 513 K water, but irradiation below 473 K did not induce IASCC at 513 K. Specimens irradiated at 333 K did not show IASCC susceptibility in 513 K water, but high susceptibility was observed in 573 K water. Effect of irradiation temperature is discussed from the view points of microstructural and microcompositional changes.

  9. Mechanical strength of martensitic 10%-Cr-steel after low-dose irradiation in HFR

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E. [Inst. fuer Materialforschung 1, Kernforschungszentrum Karlsruhe GmbH (Germany); Romer, O. [Hauptabteilung fuer Versuchstechnik/Heisse Zellen, Kernforschungszentrum Karlsruhe GmbH (Germany)

    1995-12-31

    Within the framework of the SIENA irradiation program (Steel Irradiation in an Enhanced Neutron Arrangement), the materials MANET I (DIN 1.4914) and AISI 316 L favored for the NET (Next European Torus) fusion reactor were investigated. The martensitic 9-12% chromium steels are considered as alternative materials for components of fusion reactors, because of their low He embrittlement and the good swelling behavior. Irradiations of the martensitic tensile specimens were performed in the reactor at 300 C, 400 C and 475 C, respectively with irradiation doses of 5, 10 and 15 dpa attained. Following the post-irradiation tensile tests, considerable hardening of the material was observed at low irradiation and test temperatures. In the microstructure, dislocation loops and He bubbles were found to occur as irradiation induced material changes. The dislocation loops contribute significantly to material embrittlement. (orig.).

  10. Fractal characteristics of fracture morphology of steels irradiated with high-energy ions

    Energy Technology Data Exchange (ETDEWEB)

    Xian, Yongqiang; Liu, Juan [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); University of Chinese Academy of Science, Beijing 100049 (China); Zhang, Chonghong, E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China); Chen, Jiachao [Paul Scherrer Institute, Villigen PSI (Switzerland); Yang, Yitao; Zhang, Liqing; Song, Yin [Institute of Modern Physics, Chinese Academy of Science, Lanzhou 730000 (China)

    2015-06-15

    Highlights: • Fractal dimensions of fracture surfaces of steels before and after irradiation were calculated. • Fractal dimension can effectively describe change of fracture surfaces induced by irradiation. • Correlation of change of fractal dimension with embrittlement of irradiated steels is discussed. - Abstract: A fractal analysis of fracture surfaces of steels (a ferritic/martensitic steel and an oxide-dispersion-strengthened ferritic steel) before and after the irradiation with high-energy ions is presented. Fracture surfaces were acquired from a tensile test and a small-ball punch test (SP). Digital images of the fracture surfaces obtained from scanning electron microscopy (SEM) were used to calculate the fractal dimension (FD) by using the pixel covering method. Boundary of binary image and fractal dimension were determined with a MATLAB program. The results indicate that fractal dimension can be an effective parameter to describe the characteristics of fracture surfaces before and after irradiation. The rougher the fracture surface, the larger the fractal dimension. Correlation of the change of fractal dimension with the embrittlement of the irradiated steels is discussed.

  11. Irradiation Effects in Fortiweld Steel Containing Different Boron Isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M.

    1967-07-15

    Tensile specimens and miniature impact specimens of the low alloyed pressure vessel steel Fortiweld have been irradiated at 265 deg C in R2 to two neutron doses, 6.5 x 10{sup 18} n/cm{sup 2} (> 1 MeV) and 4 x 10{sup 19} n/cm{sup 2} (thermal) and also 9.0 x 10{sup 18} n/cm{sup 2} (> 1 MeV) and 6 x 10{sup 19} n/cm{sup 2} (thermal). Material from three laboratory melts, in which the boron consisted of {sup 10}B, {sup 11}B and natural boron respectively, were investigated. The results both of tensile tests and impact tests with miniature impact specimens show that the {sup 10}B-alloyed material was changed more and the {sup 11}B-alloyed material was changed less than the material containing natural boron. At the higher neutron dose the increase in yield strength (0.2 % offset yield strength) was 11 kg/mm in the {sup 10}B containing material compared to 5 kg/mm in the {sup 11}B-containing material. The decrease in total elongation was 5 and 0 percentage units respectively. The transition temperature was increased 190 deg C at the higher neutron dose in the {sup 10}B-alloyed material, 40 deg C in the {sup 11}B-alloyed material and 80 deg C in the material containing natural boron.

  12. Embrittlement of Cr-Mo steels after low fluence irradiation in HFIR

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

    The goal of this work is the determination of the possible effect of the simultaneous formation of helium and displacement damage during irradiation on the Charpy impact behavior. Subsize Charpy impact specimens of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and 12Cr-1MoVW with 2%Ni (12Cr-1MOVW-2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400{degree}C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toghness. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr-1MoVW-2Ni steel irradiated at 400{degree}C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behaviour of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

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

    Science.gov (United States)

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

    2016-04-01

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

  14. Fracture properties of neutron-irradiated martensitic 9Cr-WVTa steels below room temperature

    Science.gov (United States)

    Abe, F.; Narui, M.; Kayano, H.

    1994-09-01

    Fracture properties of the reduced activation martensitic 9Cr-1WVTa and 9Cr-3WVTa steels were investigated by carrying out instrumented Charpy impact tests and tensile tests at temperatures below room temperature after irradiation in the Japan Materials Testing Reactor at 493 and 538 K. Modified 9Cr-1MoVNb steel was also examined for comparison. The irradiation-induced increase in ductile-to-brittle transition temperature was 53, 26 and 40 K for the {1}/{3} size Charpy specimens of 9Cr-1WVTa, 9Cr-3WVTa and 9Cr-1MoVNb steels, respectively, which resulted primarily from the irradiation-induced increase in yield stress. The cleavage fracture stress was 1820-1870 MPa for the three steels in unirradiated conditions, which was scarcely affected by irradiation. The deflections to the maximum load and to the brittle fracture initiation were decreased by irradiation. In the tensile test, quasi-cleavage fracture occurred at 77 K in both unirradiated and irradiated conditions. The cleavage fracture stress was 1320-1380 MPa for the tensile specimens of the three steels, which was about 1.4 times smaller than that for the Charpy specimens.

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

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

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

  16. Damage behavior in helium-irradiated reduced-activation martensitic steels at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-12-15

    Dislocation loops induced by helium irradiation at elevated temperatures in reduced-activation martensitic steels were investigated using transmission electron microscopy. Steels were irradiated with 100 keV helium ions to 0.8 dpa between 300 K and 723 K. At irradiation temperatures T{sub irr} ⩽ 573 K, small defects with both Burger vectors b = 1/2〈1 1 1〉 and b = 〈1 0 0〉 were observed, while at T{sub irr} ⩾ 623 K, the microstructure was dominated by large convoluted interstitial dislocation loops with b = 〈1 0 0〉. Only small cavities were found in the steels irradiated at 723 K.

  17. Deformation behavior in reactor pressure vessel steels as a clue to understanding irradiation hardening.

    Energy Technology Data Exchange (ETDEWEB)

    DiMelfi, R. J.; Alexander, D. E.; Rehn, L. E.

    1999-10-25

    In this paper, we examine the post-yield true stress vs true strain behavior of irradiated pressure vessel steels and iron-based alloys to reveal differences in strain-hardening behavior associated with different irradiating particles (neutrons and electrons) and different alloy chernky. It is important to understand the effects on mechanical properties caused by displacement producing radiation of nuclear reactor pressure steels. Critical embrittling effects, e.g. increases in the ductile-to-brittle-transition-temperature, are associated with irradiation-induced increases in yield strength. In addition, fatigue-life and loading-rate effects on fracture can be related to the post-irradiation strain-hardening behavior of the steels. All of these properties affect the expected service life of nuclear reactor pressure vessels. We address the characteristics of two general strengthening effects that we believe are relevant to the differing defect cluster characters produced by neutrons and electrons in four different alloys: two pressure vessel steels, A212B and A350, and two binary alloys, Fe-0.28 wt%Cu and Fe-0.74 wt%Ni. Our results show that there are differences in the post-irradiation mechanical behavior for the two kinds of irradiation and that the differences are related both to differences in damage produced and alloy chemistry. We find that while electron and neutron irradiations (at T {le} 60 C) of pressure vessel steels and binary iron-based model alloys produce similar increases in yield strength for the same dose level, they do not result in the same post-yield hardening behavior. For neutron irradiation, the true stress flow curves of the irradiated material can be made to superimpose on that of the unirradiated material, when the former are shifted appropriately along the strain axis. This behavior suggests that neutron irradiation hardening has the same effect as strain hardening for all of the materials analyzed. For electron irradiated steels, the

  18. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  19. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  20. 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.

  1. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  2. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  3. Heavy-section steel irradiation program. Progress report, October 1994--March 1995

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-10-01

    This document is the October 1994-March 1995 Progress Report for the Heavy Section Steel Irradiation Program. The report contains a summary of activities in each of the 14 tasks of the HSSI Program, including: (1) Program management, (2) Fracture toughness shifts in high-copper weldments, (3) Fracture toughness shifts in low upper-shelf welds, (4) Irradiation effects in a commercial low upper-shelf weld, (5) Irradiation effects on weld heat-affected zone and plate materials, (6) Annealing effects in low upper-shelf welds, (7) Microstructural analysis of radiation effects, (8) In-service irradiated and aged material evaluations, (9) Japanese power development reactor vessel steel examination, (10) fracture toughness curve shift method, (11) Special technical assistance, (12) Technical assistance for JCCCNRS, (13) Correlation monitor materials, and (14) Test reactor irradiation coordination. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  4. Modeling of helium bubble nucleation and growth in neutron irradiated boron doped RAFM steels

    Energy Technology Data Exchange (ETDEWEB)

    Dethloff, Christian, E-mail: christian.dethloff@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Gaganidze, Ermile [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Svetukhin, Vyacheslav V. [Ulyanovsk State University, Leo Tolstoy Str. 42, 432970 Ulyanovsk (Russian Federation); Aktaa, Jarir [Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-15

    Reduced activation ferritic/martensitic (RAFM) steels are promising candidates for structural materials in future fusion technology. In addition to other irradiation defects, the transmuted helium is believed to strongly influence material hardening and embrittlement behavior. A phenomenological model based on kinetic rate equations is developed to describe homogeneous nucleation and growth of helium bubbles in neutron irradiated RAFM steels. The model is adapted to different {sup 10}B doped EUROFER97 based heats, which already had been studied in past irradiation experiments. Simulations yield bubble size distributions, whereby effects of helium generation rate, surface energy, helium sinks and helium density are investigated. Peak bubble diameters under different conditions are compared to preliminary microstructural results on irradiated specimens. Helium induced hardening was calculated by applying the Dispersed Barrier Hardening model to simulated cluster size distributions. Quantitative microstructural investigations of unirradiated and irradiated specimens will be used to support and verify the model.

  5. Mechanical Test on Irradiated Welding X80/X02 Steel

    Institute of Scientific and Technical Information of China (English)

    LIU; Xin-peng; ZHANG; Chang-yi; NING; Guang-sheng; TONG; Zhen-feng; YANG; Wen

    2015-01-01

    The dedicated X80base metal,welding metal and X80/X02HAZ metal are irradiated in experimental reactor in order to evaluate the mechanical properties on the special condition.The cumulative irradiate dose(E>1 MeV)is 4×1016 cm-2,and irradiating temperature is below

  6. 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.

  7. 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).

  8. Void formation and microstructural development in oxide dispersion strengthened ferritic steels during electron-irradiation

    Science.gov (United States)

    Saito, J.; Suda, T.; Yamashita, S.; Ohnuki, S.; Takahashi, H.; Akasaka, N.; Nishida, M.; Ukai, S.

    1998-10-01

    ODS ferritic steels (13Cr-0.5Ti-0.2Y 2O 3) were prepared by the mechanical alloying method followed by the hot extrusion and several heat treatments including recrystallization. ODS steels with different heat treatment and a ferritic/martensitic (F/M) steel for the reference were irradiated to 12 dpa at 670-770 K in HVEM. After recrystallization, the dislocation density decreased with increasing grain size, however, the oxide particles did not show any obvious change in the size and the number density. During the electron-irradiation the microstructure was relatively stable, i.e. oxide particles showed good stability and the dislocation density remained almost constant. A limited void formation was observed in the specimens, and the suppressive effect due to dislocations with high number density was confirmed. From these results, the behavior of microstructure and the limited void formation in ODS steels have been discussed.

  9. Irradiation embrittlement of reactor pressure vessel steel outside the astm specification A508 CL2

    Science.gov (United States)

    Pachur, D.; Krawczynski, S. J.; Derz, H.; Pott, G.

    1990-04-01

    Radiation embrittlement of reactor pressure vessel steels is of considerable significance for safety engineering. Steel manufacturers must therefore comply with specifications defined by national design codes. The extent to which a steel deviating from the specification is influenced by irradiation is being examined under the German Research Programme on the Integrity of Reactor Components. Charpy-V specimens were taken from a forged steel block longitudinally and vertically to the direction of main deformation and irradiated in the FRJ-1 research reactor at a temperature of 288 °C corresponding to the operating temperature of power reactors. The neutron fluences obtained ranged between 0.8 × 10 19 and 8 × 10 19n/ cm2. Instrumented pendulum impact tests have been evaluated and the load signals measured were analysed, fitting and calculating transition temperature curves and trend curves.

  10. Fracture toughness of irradiated modified 9Cr-lMo steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Yoon, J.H.; Ryu, W.S.; Lee, C.B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Hong, J.H. [KAERI - Korea Atomic Energy Research Institute, Nuclear Materials Technology Development Div., Daejon (Korea, Republic of)

    2007-07-01

    Full text of publication follows: Ferritic/martensitic steels have been used for a long time in the power generation industry as boiler and turbine materials. These steels are the proposed candidates for the crosscutting materials of the advanced nuclear power system. It is important to realize the change of mechanical properties by neutron irradiation for application these materials to nuclear power system. Irradiation effect on the fracture toughness of the structural materials is one of the concerns for the designing of the fusion devices. The test material was a 16 mm thick commercial Modified 9Cr-1Mo plate which was normalized at 1050 deg. C and tempered at 770 deg. C. The half sized pre-cracked Charpy specimens were irradiated at CT test hole in HANARO. Irradiation test was conducted at 340 deg. C and 400 deg. C to investigate the irradiation temperature effect on the degradation of the fracture toughness. And the irradiation fluence was 1.2x10{sup 21} n/cm{sup 2} (E>0.1 MeV). Toughness tests for the irradiated specimens will be performed in the hot cell at KAERI. The fracture toughness of the unirradiated condition was carried out in order to assess the changes in the materials properties caused by neutron irradiation. The K{sub JC} values in accordance at the ASTM E1921- 05 standard were obtained by three-point bending tests. Tests have been carried out at several temperatures within transition region. The multi-temperature method was used to determine reference temperature, T{sub o}. The applicability of the Master Curve method for irradiated and unirradiated ferritic/martensitic steel is another focus of this study. The reference temperature of the unirradiated specimen was -72.4 deg. C. And the Master Curve successfully expressed the trend of the fracture toughness change with temperature for unirradiated Modified 9Cr-1Mo steel. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kohyama, Akira [Kyoto Univ., Institute of Advanced Energy, Uji, Kyoto (Japan); Donomae, Takako [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    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)

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Z. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)], E-mail: zheng.lu@lboro.ac.uk; Faulkner, R.G.; Morgan, T.S. [IPTME, Loughborough University, Loughborough LE11 3U (United Kingdom)

    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 x 10{sup -6} dpa/s) at 400 deg. C and 28 dpa (1.7 x 10{sup -6} dpa/s) at 465 deg. 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 ({approx}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.

  14. Positron annihilation Doppler broadening spectroscopy study on Fe-ion irradiated NHS steel

    Science.gov (United States)

    Zhu, Huiping; Wang, Zhiguang; Gao, Xing; Cui, Minghuan; Li, Bingsheng; Sun, Jianrong; Yao, Cunfeng; Wei, Kongfang; Shen, Tielong; Pang, Lilong; Zhu, Yabin; Li, Yuanfei; Wang, Ji; Song, Peng; Zhang, Peng; Cao, Xingzhong

    2015-02-01

    In order to study the evolution of irradiation-induced vacancy-type defects at different irradiation fluences and temperatures, a new type of ferritic/martensitic (F/M) steel named NHS (Novel High Silicon) was irradiated by 3.25 MeV Fe-ion at room temperature and 723 K to fluences of 4.3 × 1015 and 1.7 × 1016 ions/cm2. After irradiation, vacancy-type defects were investigated with variable-energy positron beam Doppler broadening spectra. Energetic Fe-ions produced a large number of vacancy-type defects in the NHS steel, but one single main type of vacancy-type defect was observed in both unirradiated and irradiated samples. The concentration of vacancy-type defects decreased with increasing temperature. With the increase of irradiation fluence, the concentration of vacancy-type defects increased in the sample irradiated at RT, whereas for the sample irradiated at 723 K, it decreased. The enhanced recombination between vacancies and excess interstitial Fe atoms from deeper layers, and high diffusion rate of self-interstitial atoms further improved by diffusion via grain boundary and dislocations at high temperature, are thought to be the main reasons for the reversed trend of vacancy-type defects between the samples irradiated at RT and 723 K.

  15. Impact behavior of reduced-activation steels irradiated to 24 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Laboratory, TN (United States)

    1996-04-01

    Charpy impact properties of eight reduced-activation Cr-W ferritic steels were determined after irradiation to {approx}21-24 dpa in the Fast Flux Test Facility (FFTF) at 365{degree}C. Chromium concentrations in the eight steels ranged from 2.25 to 12wt% Cr (steels contained {approx}0.1%C). the 2 1/4Cr steels contained variations of tungsten and vanadium, and the steels with 5, 9, and 12% Cr, contained a combination of 2% W and 0.25% V. A 9Cr in FFTF to {approx}6-8 and {approx}15-17 dpa. Irradiation caused an increase in the DBTT and decrease in the USE, but there was little further change in the DBTT from that observed after the 15-17 dpa irradiation, indicating that the shift had essentially saturated with fluence. The results are encouraging because they indicate that the effect of irradiation on toughness can be faorably affected by changing composition and microstructure.

  16. Thermally activated deformation of irradiated reactor pressure vessel steel

    Science.gov (United States)

    Böhmert, J.; Müller, G.

    2002-03-01

    Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

  17. Microstructure evolution and degradation mechanisms of reactor internal steel irradiated with heavy ions

    Science.gov (United States)

    Borodin, O. V.; Bryk, V. V.; Kalchenko, A. S.; Parkhomenko, A. A.; Shilyaev, B. A.; Tolstolutskaya, G. D.; Voyevodin, V. N.

    2009-03-01

    Structure evolution and degradation mechanisms during irradiation of 18Cr-10Ni-Ti steel (material of VVER-1000 reactor internals are investigated). Using accelerator irradiations with Cr3+ and Ar+ ions allowed studying effects of dose rate, different initial structure state and implanted ions on features of structure evolution and main mechanisms of degradation including low temperature swelling and embrittlement of the 18Cr-10Ni-Ti steel. It is shown that differences in dose rate at most irradiation temperatures mainly exert their influence on the duration of the swelling transient regime. Calculations of possible transmutation products during irradiation of this steel in a VVER-1000 spectrum were performed. It is shown that gaseous atoms (He and H), which are generated simultaneously with radiation defects, stabilize the elements of radiation microstructure and influence the swelling. The nature of deformation under different temperatures of irradiation and of mechanical testing is investigated. It is shown that the temperature sensitivity of swelling behaviour in the investigated steel, with different initial structures can be connected with the dynamic behaviour of point defect sinks.

  18. Superior Charpy impact properties of ODS ferritic steel irradiated in JOYO

    Science.gov (United States)

    Kuwabara, T.; Kurishita, H.; Ukai, S.; Narui, M.; Mizuta, S.; Yamazaki, M.; Kayano, H.

    1998-10-01

    The effect of neutron irradiation on Charpy impact properties of an ODS ferritic steel developed by PNC was studied. The miniaturized Charpy V-notch (MCVN) specimens (1.5 × 1.5 × 20 mm) of two orientations (longitudinal, called 1DS-L, and transverse, 1DS-T) were irradiated to fluence levels of (0.3-3.8) × 10 26 n/m 2 ( E n > 0.1 MeV) between 646 and 845 K in JOYO. MCVN specimens before and after the irradiation were subjected to instrumented Charpy impact tests. The test results and fracture surface observations showed that in the unirradiated state the steel showed no ductile-to-brittle transition behavior until 153 K regardless of orientation and the upper shelf energy of the steel was as high as that of a high-strength ferritic steel without dispersed oxide. Such excellent impact properties were essentially maintained after the irradiation although an appreciable decrease in absorbed energy occurred by higher temperature irradiations at and above 793 K.

  19. Positron annihilation Doppler broadening spectroscopy study on Fe-ion irradiated NHS steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Huiping [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Wang, Zhiguang, E-mail: zhgwang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Gao, Xing; Cui, Minghuan; Li, Bingsheng; Sun, Jianrong; Yao, Cunfeng; Wei, Kongfang; Shen, Tielong; Pang, Lilong; Zhu, Yabin [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Li, Yuanfei; Wang, Ji [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); University of Lanzhou, Lanzhou 730000 (China); Song, Peng [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Peng; Cao, Xingzhong [Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049 (China)

    2015-02-01

    Highlights: • NHS steel was irradiated by 3.25 MeV Fe ions to different fluences at room temperature and 723 K. • The evolution of vacancy type defects is studied through positron annihilation spectroscopy. • The concentration gradient of SIA can induce the decrease of S value with increasing fluence at high temperature. - Abstract: In order to study the evolution of irradiation-induced vacancy-type defects at different irradiation fluences and temperatures, a new type of ferritic/martensitic (F/M) steel named NHS (Novel High Silicon) was irradiated by 3.25 MeV Fe-ion at room temperature and 723 K to fluences of 4.3 × 10{sup 15} and 1.7 × 10{sup 16} ions/cm{sup 2}. After irradiation, vacancy-type defects were investigated with variable-energy positron beam Doppler broadening spectra. Energetic Fe-ions produced a large number of vacancy-type defects in the NHS steel, but one single main type of vacancy-type defect was observed in both unirradiated and irradiated samples. The concentration of vacancy-type defects decreased with increasing temperature. With the increase of irradiation fluence, the concentration of vacancy-type defects increased in the sample irradiated at RT, whereas for the sample irradiated at 723 K, it decreased. The enhanced recombination between vacancies and excess interstitial Fe atoms from deeper layers, and high diffusion rate of self-interstitial atoms further improved by diffusion via grain boundary and dislocations at high temperature, are thought to be the main reasons for the reversed trend of vacancy-type defects between the samples irradiated at RT and 723 K.

  20. Effect of irradiation on the steels 316L/LN type to 12 dpa at 400 °C

    Science.gov (United States)

    Bulanova, T.; Fedoseev, A.; Kalinin, G.; Rodchenkov, B.; Shamardin, V.

    2004-08-01

    The 316L type stainless steel is widely used as a structural material for the fission reactors internal structures (core, core supports, etc.) and for experimental irradiation facilities. The 316L(N)-IG type steel is proposed as a main structural material for the ITER reactor (first wall, blanket, vacuum vessel, cooling pipe lines). It is obvious that different steel grades should exhibit different reaction to neutron irradiation. The main objective of this work was to study of irradiation behaviour of three different commercial steels: AISI 316LN, AISI 316L (US grades) and 02X17H14M2 (Russian steel grade that is similar to 316L). Irradiation effect on the three commercial steels of 316L family to ˜12 dpa at the temperature ˜370-400 °C on the tensile properties, microstructure, swelling and susceptibility to SCC are described in the paper.

  1. Effect of irradiation on the steels 316L/LN type to 12 dpa at 400 deg. C

    Energy Technology Data Exchange (ETDEWEB)

    Bulanova, T. E-mail: fae@niiar.rukalinig@nikiet.ru; Fedoseev, A.; Kalinin, G.; Rodchenkov, B.; Shamardin, V

    2004-08-01

    The 316L type stainless steel is widely used as a structural material for the fission reactors internal structures (core, core supports, etc.) and for experimental irradiation facilities. The 316L(N)-IG type steel is proposed as a main structural material for the ITER reactor (first wall, blanket, vacuum vessel, cooling pipe lines). It is obvious that different steel grades should exhibit different reaction to neutron irradiation. The main objective of this work was to study of irradiation behaviour of three different commercial steels: AISI 316LN, AISI 316L (US grades) and 02X17H14M2 (Russian steel grade that is similar to 316L). Irradiation effect on the three commercial steels of 316L family to {approx}12 dpa at the temperature {approx}370-400 deg. C on the tensile properties, microstructure, swelling and susceptibility to SCC are described in the paper.

  2. Evolution of microstructure after irradiation creep in several austenitic steels irradiated up to 120 dpa at 320 °C

    Energy Technology Data Exchange (ETDEWEB)

    Renault-Laborne, A., E-mail: alexandra.renault@cea.fr [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Garnier, J.; Malaplate, J. [DEN-Service de Recherches Métallurgiques Appliquées, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Gavoille, P. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Sefta, F. [EDF R& D, MMC, Site des Renardières, F-77818, Morêt-sur-Loing Cedex (France); Tanguy, B. [DEN-Service d' Etudes des Matériaux Irradiés, CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2016-07-15

    Irradiation creep was investigated in different austenitic steels. Pressurized tubes with stresses of 127–220 MPa were irradiated in BOR-60 at 320 °C to 120 dpa. Creep behavior was dependent on both chemical composition and metallurgical state of steels. Different steels irradiated with and without stress were examined by TEM. Without stress, the irradiation produced high densities of dislocation lines and Frank loops and, depending on the type of steels, precipitates. Stress induced an increase of the precipitate mean size and density and, for some grades, an increase of the mean loop size and a decrease of their density. An anisotropy of Frank loop density or size induced by stress was not observed systematically. Dislocation line microstructure seems not to be different between the stressed and unstressed specimens. No cavities were detectable in these specimens. By comparing with the data from this work, the main irradiation creep models are discussed.

  3. The microstructure of neutron irradiated type-348 stainless steel and its relation to creep and hardening

    Science.gov (United States)

    Thomas, L. E.; Beeston, J. M.

    1982-06-01

    Annealed type-348 stainless steel specimens irradiated to 33 to 39 dpa at 350°C were examined by transmission electron microscopy to determine the cause of pronounced irradiation creep and hardening. The irradiation produced very high densities of 1-2 nm diameter helium bubbles, 2-20 nm diameter faulted (Frank) dislocation loops and 10 nm diameter precipitate particles. These defects account for the observed irradiation hardening but do not explain the creep strains. Too few point defects survive as faulted dislocation loops for significant creep by the stress-induced preferential absorption (SIPA) mechanism and there are not enough unfaulted dislocations for creep by climb-induced glide. Also, the irradiation-induced precipitates are face-centred cubic G-phase (a niobium nickel suicide), and cannot cause creep. It is suggested that the irradiation creep occurs by a grain-boundary movement mechanism such as diffusion accomodated grain-boundary sliding.

  4. 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.

  5. 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.

  6. Impact of the nanostructuration on the corrosion resistance and hardness of irradiated 316 austenitic stainless steels

    Science.gov (United States)

    Hug, E.; Prasath Babu, R.; Monnet, I.; Etienne, A.; Moisy, F.; Pralong, V.; Enikeev, N.; Abramova, M.; Sauvage, X.; Radiguet, B.

    2017-01-01

    The influence of grain size and irradiation defects on the mechanical behavior and the corrosion resistance of a 316 stainless steel have been investigated. Nanostructured samples were obtained by severe plastic deformation using high pressure torsion. Both coarse grain and nanostructured samples were irradiated with 10 MeV 56Fe5+ ions. Microstructures were characterized using transmission electron microscopy and atom probe tomography. Surface mechanical properties were evaluated thanks to hardness measurements and the corrosion resistance was studied in chloride environment. Nanostructuration by high pressure torsion followed by annealing leads to enrichment in chromium at grain boundaries. However, irradiation of nanostructured samples implies a chromium depletion of the same order than depicted in coarse grain specimens but without metallurgical damage like segregated dislocation loops or clusters. Potentiodynamic polarization tests highlight a definitive deterioration of the corrosion resistance of coarse grain steel with irradiation. Downsizing the grain to a few hundred of nanometers enhances the corrosion resistance of irradiated samples, despite the fact that the hardness of nanocrystalline austenitic steel is only weakly affected by irradiation. These new experimental results are discussed in the basis of couplings between mechanical and electrical properties of the passivated layer thanks to impedance spectroscopy measurements, hardness properties of the surfaces and local microstructure evolutions.

  7. Damage structure of austenitic stainless steel 316LN irradiated at low temperature in HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, N.; Robertson, J.P.; Grossbeck, M.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Wakai, E. [Japan Atomic Energy Research Inst. (Japan)

    1998-03-01

    TEM disk specimens of austenitic stainless steel 316LN irradiated to damage levels of about 3 dpa at irradiation temperatures of either about 90 C or 250 C have been investigated by using transmission electron microscopy. The irradiation at 90 C and 250 C induced a dislocation loop density of 3.5 {times} 10{sup 22} m{sup {minus}3} and 6.5 {times} 10{sup 22} m{sup {minus}3}, a black dot density of 2.2 {times} 10{sup 23} m{sup {minus}3} and 1.6 {times} 10{sup 23} m{sup {minus}3}, respectively, in the steels, and a high density (<1 {times} 10{sup 22} m{sup {minus}3}) of precipitates in matrix. Cavities could be observed in the specimens after the irradiation. It is suggested that the dislocation loops, the black dots, and the precipitates cause irradiation hardening, an increase in the yield strength and a decrease in the uniform elongation, in the 316LN steel irradiated at low temperature.

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

    Science.gov (United States)

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

    2015-11-01

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

  9. Irradiation response of ODS ferritic steels to high-energy Ne ions at HIRFL

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, C.H., E-mail: c.h.zhang@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou (China); Yang, Y.T.; Song, Y. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou (China); Chen, J. [Paul Scherrer Institut, Villigen PSI (Switzerland); Zhang, L.Q. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou (China); Jang, J. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kimura, A. [Institute of Advanced Energy, Kyoto University, Kyoto (Japan)

    2014-12-15

    Two kinds of ODS high-Cr ferritic steels (commercial MA956 and an Al-free 16Cr–0.1Ti ODS ferritic steel) and one conventional ferritic/martensitic steel (T122) were irradiated at about 440 °C with high-energy {sup 20}Ne-ions in HIRFL. Successively increasing doses from 350 to 900 appm of Ne concentration, corresponding to atomic displacement levels from 0.7 to 1.8 dpa, were approached. A nearly uniform distribution of Ne concentration and atomic displacement damage was produced through the thickness of 60 μm of the specimens by using an energy degrader. Mechanical properties of the specimens were tested with the small-ball punch technique. The test at room temperature shows a less significant ductility loss in the ODS ferritic steel MA956 than in the T122 irradiated to the same dose of 350 appm Ne/0.7 dpa. The test at 500 °C shows that the Al-free 16Cr–0.1Ti ODS ferritic steel does not exhibit observable loss of ductility even to the highest dose level (900 appm Ne/1.8 dpa). An investigation with transmission electron microscopy (TEM) shows that voids with a diameter up to 70 nm were formed at grain boundaries in the conventional ferritic/martensitic steel T122 while only smaller bubbles were formed at the oxides/substrate interfaces in the ODS ferritic steel MA956. Mechanisms underlying the difference of irradiation response of the steels are discussed.

  10. Radiation hardening and deformation behavior of irradiated ferritic-martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, J.P.; Klueh, R.L.; Rowcliffe, A.F. [Oak Ridge National Lab., TN (United States); Shiba, K. [Japan Atomic Energy Research Inst. (Japan)

    1998-03-01

    Tensile data from several 8--12% Cr alloys irradiated in the High Flux Isotope Reactor (HFIR) to doses up to 34 dpa at temperatures ranging from 90 to 600 C are discussed in this paper. One of the critical questions surrounding the use of ferritic-martensitic steels in a fusion environment concerns the loss of uniform elongation after irradiation at low temperatures. Irradiation and testing at temperatures below 200--300 C results in uniform elongations less than 1% and stress-strain curves in which plastic instability immediately follows yielding, implying dislocation channeling and flow localization. Reductions in area and total elongations, however, remain high.

  11. Initial tensile test results from J316 stainless steel irradiated in the HFIR spectrally tailored experiment

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

    The objective of this work is to determine the effects of neutron irradiation on the mechanical properties of austenitic stainless steel alloys. In this experiment, the spectrum has been tailored to reduce the thermal neutron flux and achieve a He/dpa level near that expected in a fusion reactor.

  12. The Early Characterization of Irradiation Effects in Stainless Steels at the Experimental Breeder Reactor-II

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter

    2008-01-01

    The new Global Nuclear Energy Partnership (GNEP) program is revitalizing interest in materials development for fast spectrum reactors. With this comes the need for new, high-performance materials that are resistant to property changes caused by radiation damage. In the 1970s there was an effort to monitor the irradiation effects on stainless steels used in fast reactor cores, largely because there were a number of ‘surprises’ where materials subjected to a high flux of fast neutrons incurred dimensional and property changes that had not been expected. In the U.S., this applied to the Experimental Breeder Reactor-II. Void swelling and irradiation-induced creep caused dimensional changes in the reactor components that shortened their useful lifetime and impacted reactor operations by creating fuel handling difficulties and reactivity anomalies. The surveillance programs and early experiments studied the simplest of austenitic stainless steels, such as Types 304 and 304L stainless steel, and led to some basic understanding of the links between these irradiation effects and microchemical changes within the steel caused by operational variables such as temperature, neutron flux and neutron fluence. Some of the observations helped to define later alloy development programs designed to produce alloys that were much more resistant to the effects of neutron irradiation.

  13. Structural Transformations in Austenitic Stainless Steel Induced by Deuterium Implantation: Irradiation at 295 K

    National Research Council Canada - National Science Library

    Morozov, Oleksandr; Zhurba, Volodymir; Neklyudov, Ivan; Mats, Oleksandr; Progolaieva, Viktoria; Boshko, Valerian

    2016-01-01

    ...—the linear region of high implantation doses (8 × 1017 to 2.7 × 1018 D/cm2). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps...

  14. Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

    Institute of Scientific and Technical Information of China (English)

    T. Hino; K. Yamaguchi; Y. Yamauchi; Y. Hirohata; K. Tsuzuki; Y.Kusama

    2005-01-01

    Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall,vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study,deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus.After a ferritic steel sample was irradiated by 1.7 kev D+ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fluence dependence of retained amount of deuterium was measured by changing the ion fluence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316 L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention.Ferritic steel sample was

  15. Radiation Response of a 9 Cr Oxide Dispersion Strengthened Steel to Heavy Ion Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Allen, Todd R. [University of Wisconsin, Madison; Gan, J. [Idaho National Laboratory (INL); Cole, James I. [Idaho National Laboratory (INL); Miller, Michael K [ORNL; Busby, Jeremy T [ORNL; Ukai, S. [Japan Atomic Energy Agency (JAEA); Shutthanandan, S. [Pacific Northwest National Laboratory (PNNL); Thevuthasan, S. [Pacific Northwest National Laboratory (PNNL)

    2008-01-01

    Ferritic-martensitic (FM) alloys are expected to play an important role as cladding or structural components in Generation IV systems operating in the temperature range 350-700 C and to doses up to 200 dpa. Oxide dispersion strengthened (ODS) ferritic-martensitic steels have been developed to operate at higher temperatures than traditional FM steels. These steels contain nanometer-sized Y-Ti-O nanoclusters as a strengthening mechanism. Heavy ion irradiation has been used to determine the nanocluster stability over a temperature range of 500-700 C to doses of 150 dpa. At all temperatures, the average nanocluster size decreases but the nanocluster density increases. The increased density of smaller nanoclusters under radiation should lead to strengthening of the matrix. While a reduction in size under irradiation has been reported in some other studies, many report oxide stability. The data from this study are contrasted to the available literature to highlight the differences in the reported radiation response.

  16. DISCUSSION ON DEFECTS DISTRIBUTION NEAR THE STEEL SURFACE IRRADIATED BY INTENSE PULSED ION BEAM

    Institute of Scientific and Technical Information of China (English)

    X.Y.Le; S.Yan; W.J.Zhao; B.X.Han; W.Xiang

    2002-01-01

    The surface defect distribution in stainless steel irradiated with intense pulsed ion beam(IPIB) of current density above 60A/cm2 and acceleration voltage 300-500keV wasdiscussed and analyzed. The defects near the surface of stainless steel were generatedin two ways: (1) generated by the bombardment of energetic ions and (2) induced bythe high level stress near the surface. Thus the temperature and stress distributionsnear the steel surface were calculated by means of our STEIPIB code, which treatedwith the thermal-dynamical process in the target irradiated by the IPIB. Based onthese distributions, the generations and movements of these defects were discussedand compared with the experiment results.

  17. Irradiation creep in austenitic and ferritic steels irradiated in a tailored neutron spectrum to induce fusion reactor levels of helium

    Energy Technology Data Exchange (ETDEWEB)

    Grossbeck, M.L.; Gibson, L.T. [Oak Ridge National Laboratory, TN (United States); Jitsukawa, S.

    1996-04-01

    Six austenitic stainless steels and two ferritic alloys were irradiated sequentially in two research reactors where the neutron spectrum was tailored to produce a He production rate typical of a fusion device. Irradiation began in the Oak Ridge Research Reactor where an atomic displacement level of 7.4 dpa was achieved and was then transferred to the High Flux Isotope Reactor for the remainder of the irradiation to a total displacement level of 19 dpa. Temperatures of 60 and 330{degree}C are reported on. At 330{degree}C irradiation creep was found to be linear in stress and fluence with rates in the range of 1.7 - 5.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. Annealed and cold-worked materials exhibited similar creep rates. There is some indication that austenitic alloys with TiC or TiO precipitates had a slightly higher irradiation creep rate than those without. The ferritic alloys HT-9 and Fe-16Cr had irradiatoin creep rates about 0.5 x 10{sup -4}% MPa{sup -1} dpa{sup -1}. No meaningful data could be obtained from the tubes irradiated at 60{degree}C because of damage to the tubes.

  18. Recent status and improvement of reduced-activation ferritic-martensitic steels for high-temperature service

    Science.gov (United States)

    Tan, L.; Katoh, Y.; Tavassoli, A.-A. F.; Henry, J.; Rieth, M.; Sakasegawa, H.; Tanigawa, H.; Huang, Q.

    2016-10-01

    Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. In addition to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M23C6 (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. Limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic are presented and compared with data for F82H and Eurofer97 irradiated up to ∼70 displacements per atom at ∼300-325 °C.

  19. Crack growth rates and fracture toughness of irradiated austenitic stainless steels in BWR environments.

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-01-21

    In light water reactors, austenitic stainless steels (SSs) are used extensively as structural alloys in reactor core internal components because of their high strength, ductility, and fracture toughness. However, exposure to high levels of neutron irradiation for extended periods degrades the fracture properties of these steels by changing the material microstructure (e.g., radiation hardening) and microchemistry (e.g., radiation-induced segregation). Experimental data are presented on the fracture toughness and crack growth rates (CGRs) of wrought and cast austenitic SSs, including weld heat-affected-zone materials, that were irradiated to fluence levels as high as {approx} 2x 10{sup 21} n/cm{sup 2} (E > 1 MeV) ({approx} 3 dpa) in a light water reactor at 288-300 C. The results are compared with the data available in the literature. The effects of material composition, irradiation dose, and water chemistry on CGRs under cyclic and stress corrosion cracking conditions were determined. A superposition model was used to represent the cyclic CGRs of austenitic SSs. The effects of neutron irradiation on the fracture toughness of these steels, as well as the effects of material and irradiation conditions and test temperature, have been evaluated. A fracture toughness trend curve that bounds the existing data has been defined. The synergistic effects of thermal and radiation embrittlement of cast austenitic SS internal components have also been evaluated.

  20. Microstructural evolution of reduced-activation martensitic steel under single and sequential ion irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Zheng, Zhongcheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yang, Feng; Xiong, Xuesong; Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2013-07-15

    Microstructural evolution of super-clean reduced-activation martensitic steels irradiated with single-beam (Fe{sup +}) and sequential-beam (Fe{sup +} plus He{sup +}) at 350 °C and 550 °C was studied. Sequential-beam irradiation induced smaller size and larger number density of precipitates compared to single-beam irradiation at 350 °C. The largest size of cavities was observed after sequential-beam irradiation at 550 °C. The segregation of Cr and W and depletion of Fe in carbides were observed, and the maximum depletion of Fe and enrichment of Cr occurred under irradiation at 350 °C.

  1. Phase Transformations in Austenitic 0Cr18Ni10Ti Steel Irradiated with High-Energy Heavy Ions

    CERN Document Server

    Hofmann, A; Semina, V K

    2000-01-01

    Radiation-induced segregation and phase transformations in 0Cr18Ni10Ti steel irradiated with high-energy heavy Ar^{+6} ions at 625^o up to 1 dpa (from 0.01 to 1 dpa) have been studied. It was found that ion irradiation accelerates carbide precipitation and EDX-analysis showed irradiation-induced segregation near grain boundaries.

  2. On the formation of stacking fault tetrahedra in irradiated austenitic stainless steels – A literature review

    Energy Technology Data Exchange (ETDEWEB)

    Schibli, Raluca, E-mail: raluca.stoenescu@gmail.com; Schäublin, Robin

    2013-11-15

    Irradiated austenitic stainless steels, because of their low stacking fault energy and high shear modulus, should exhibit a high ratio of stacking fault tetrahedra relative to the overall population of radiation induced nanometric defects. Experimental observations of stacking fault tetrahedra by transmission electron microscopy in commercial-purity stainless steels are however scarce, while they abundantly occur in high-purity or model austenitic alloys irradiated at both low and high temperatures, but not at around 673 K. In commercial alloys, the little evidence of stacking fault tetrahedra does not follow such a trend. These contradictions are reviewed and discussed. Reviewing the three possible formation mechanisms identified in the literature, namely the Silcox and Hirsch Frank loop dissociation, the void collapse and the stacking fault tetrahedra growth, it seems that the later dominates under irradiation.

  3. Void denuded zone formation for Fe–15Cr–15Ni steel and PNC316 stainless steel under neutron and electron irradiations

    Energy Technology Data Exchange (ETDEWEB)

    Sekio, Yoshihiro, E-mail: sekio.yoshihiro@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Yamashita, Shinichiro [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Sakaguchi, Norihito [Center for Advanced Research of Energy Technology, Hokkaido University, Hokkaido 060-0808 (Japan); Takahashi, Heishichiro [Oarai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Center for Advanced Research of Energy Technology, Hokkaido University, Hokkaido 060-0808 (Japan)

    2015-03-15

    Highlights: • Austenitic stainless steel developed to improve void swelling was used. • Void denuded zone formed near grain boundary can be affected by vacancy mobility. • Vacancy migration energy was estimated from void denuded zone width in the steel. - Abstract: Irradiation-induced void denuded zone (VDZ) formation near grain boundaries was studied to clarify the effects of minor alloying elements on vacancy diffusivity during irradiation in practical PNC316 stainless steel developed for nuclear reactor core materials. The test materials were Fe–15Cr–15Ni steel without additives and PNC316 stainless steel; the latter contains minor alloying elements to improve the void swelling resistance. These steels were neutron-irradiated in the experimental fast reactor JOYO at temperatures from 749 K to 775 K and fast neutron doses of 18–103 dpa, and electron irradiation was also carried out using 1 MeV high voltage electron microscopy at temperatures of 723 K and 773 K and doses up to 14.4 dpa. VDZ formation was analyzed by TEM microstructural observation after irradiation by considering radiation-induced segregation near the grain boundaries. VDZs were formed near random grain boundaries with higher misfit angles in both Fe–15Cr–15Ni and PNC316 steels. The VDZ widths in the PNC316 stainless steel were narrower than those for the Fe–15Cr–15Ni steel for all neutron and electron irradiations. The VDZ width analysis implied that the vacancy diffusivity was reduced in PNC316 stainless steel as a result of interaction of vacancies with minor alloying elements.

  4. Hardening of ODS ferritic steels under irradiation with high-energy heavy ions

    Science.gov (United States)

    Ding, Z. N.; Zhang, C. H.; Yang, Y. T.; Song, Y.; Kimura, A.; Jang, J.

    2017-09-01

    Influence of the nanoscale oxide particles on mechanical properties and irradiation resistance of oxide-dispersion-strengthened (ODS) ferritic steels is of critical importance for the use of the material in fuel cladding or blanket components in advanced nuclear reactors. In the present work, impact of structures of oxide dispersoids on the irradiation hardening of ODS ferritic steels was studied. Specimens of three high-Cr ODS ferritic steels containing oxide dispersoids with different number density and average size were irradiated with high-energy Ni ions at about -50 °C. The energy of the incident Ni ions was varied from 12.73 MeV to 357.86 MeV by using an energy degrader at the terminal so that a plateau of atomic displacement damage (∼0.8 dpa) was produced from the near surface to a depth of 24 μm in the specimens. A nanoindentor (in constant stiffness mode with a diamond Berkovich indenter) and a Vickers micro-hardness tester were used to measure the hardeness of the specimens. The Nix-Gao model taking account of the indentation size effect (ISE) was used to fit the hardness data. It is observed that the soft substrate effect (SSE) can be diminished substantially in the irradiated specimens due to the thick damaged regions produced by the Ni ions. A linear correlation between the nano-hardeness and the micro-hardness was found. It is observed that a higher number density of oxide dispersoids with a smaller average diameter corresponds to an increased resistance to irradiation hardening, which can be ascribed to the increased sink strength of oxides/matrix interfaces to point defects. The rate equation approach and the conventional hardening model were used to analyze the influence of defect clusters on irradiation hardening in ODS ferritic steels. The numerical estimates show that the hardening caused by the interstitial type dislocation loops follows a similar trend with the experiment data.

  5. High temperature deformation behavior, thermal stability and irradiation performance in Grade 92 steel

    Science.gov (United States)

    Alsagabi, Sultan

    The 9Cr-2W ferritic-martensitic steel (i.e. Grade 92 steel) possesses excellent mechanical and thermophysical properties; therefore, it has been considered to suit more challenging applications where high temperature strength and creep-rupture properties are required. The high temperature deformation mechanism was investigated through a set of tensile testing at elevated temperatures. Hence, the threshold stress concept was applied to elucidate the operating high temperature deformation mechanism. It was identified as the high temperature climb of edge dislocations due to the particle-dislocation interactions and the appropriate constitutive equation was developed. In addition, the microstructural evolution at room and elevated temperatures was investigated. For instance, the microstructural evolution under loading was more pronounced and carbide precipitation showed more coarsening tendency. The growth of these carbide precipitates, by removing W and Mo from matrix, significantly deteriorates the solid solution strengthening. The MX type carbonitrides exhibited better coarsening resistance. To better understand the thermal microstructural stability, long tempering schedules up to 1000 hours was conducted at 560, 660 and 760°C after normalizing the steel. Still, the coarsening rate of M23C 6 carbides was higher than the MX-type particles. Moreover, the Laves phase particles were detected after tempering the steel for long periods before they dissolve back into the matrix at high temperature (i.e. 720°C). The influence of the tempering temperature and time was studied for Grade 92 steel via Hollomon-Jaffe parameter. Finally, the irradiation performance of Grade 92 steel was evaluated to examine the feasibility of its eventual reactor use. To that end, Grade 92 steel was irradiated with iron (Fe2+) ions to 10, 50 and 100 dpa at 30 and 500°C. Overall, the irradiated samples showed some irradiation-induced hardening which was more noticeable at 30°C. Additionally

  6. Investigation of mechanical properties and proton irradiation behaviors of SA-738 Gr.B steel used as reactor containment

    Directory of Open Access Journals (Sweden)

    Ma Yongzheng

    2016-08-01

    Full Text Available The proton irradiation behaviors of two kinds of SA-738Gr.B steels prepared by different heat treatment used as AP1000 reactor containment were investigated by transmission electron microscopy and positron annihilation lifetime spectrum (PAS. The mechanical properties of as-received steels were also measured. In the unirradiated conditions, the SA-738Gr.B steels had high tensile strength and excellent impact fracture toughness, which met the performance requirements of ASME codes. Both kinds of SA-738Gr.B steels were irradiated by 400keV proton from 1.07×1017H+/cm2 to 5.37×1017H+/cm2 fluence at 150 ºC. Some voids and dislocation loops with several nanometers were observed in the cross-section irradiated samples prepared by electroplating and then twin-jet electropolishing technology. The number of irradiation defects increased with increasing of displacement damage, as well as for the mean positron lifetimes. The stress-relief annealing treatment improved irradiation resistance based on open volume defect analysis from proton irradiation. SA-738Gr.B (SR steel had higher proton irradiation resistance ability than that of SA-738Gr.B (QT steel. The mechanism of irradiation behaviors were also analyzed and discussed.

  7. The corrosion resistance of Eurofer 97 and ODS-Eurofer steels for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Terada, M. [Escola Politecnica da Univ. de Sao Paulo, Dept. de Engenharia Metalurgica e de Materiais, Sao Paulo-SP (Brazil); Zschommler Sandim, H.R. [Sao Paulo Univ., Dept. de Engenharia de Materiais, Polo Urbo-Industrial, Lorena-SP (Brazil); Costa, I. [Instituto de Pesquisas Energeticas e Nucleares IPEN-CCTM, Sao Paulo - SP (Brazil); Padilha, A.F. [Escola Politecnica da Universidade de Sao Paulo, Dept. de Engenharia Metalurgica e de Materiais, Sao Paulo-SP (Brazil)

    2009-07-01

    Reduced-activation-ferritic-martensitic (RAFM) steels are considered for application in fusion technology as structural materials for the first wall of future fusion reactors DEMO. Ferritic-martensitic steels show reasonably good thermo-physical and mechanical properties, low sensitivity to radiation-induced swelling and helium embrittlement under (fission) neutron irradiation and good compatibility with major cooling and breeding materials. In recent years, reduced activation versions of this type of steels have been developed in Japan and Europe in laboratory scale and tested with equivalent or even better mechanical properties. In result of a systematic development of reduced activation ferritic-martensitic (RAFM) steels in Europe, the 9% CrWVTa alloy EUROFER was specified, and industrial batches have been produced in a variety of different semi-finished product forms. The EUROFER 97 alloy was developed on the basis of the experience gained with steels of the OPTIFER, MANET and F82H-modified type. Oxide dispersion to strengthen (ODS) alloys have been used in order to increase the working temperature of RAFM steels increasing their potentiality for applications in fusion reactors that operate at temperatures higher than 650 C. The literature on the corrosion properties of these alloys is scarce. In the present work the corrosion resistance of EUROFER 97 and ODS-EUROFER was tested in solutions containing H{sub 2}SO{sub 4} and KSCN at 25 C. The results were compared to those of AISI 430 ferritic and AISI 410 martensitic conventional stainless steels. The as-received samples were tested by electrochemical techniques, specifically, potentiodynamic polarization curves and double loop electrochemical potentio-kinetic reactivation tests. The surfaces were observed by scanning electron microscopy (SEM) after exposure to corrosive media. The results showed that EUROFER 97 and ODS-EUROFER present similar corrosion resistance but lower than that of ferritic AISI 430 and

  8. Chromium redistribution in thermally aged and irradiated ferritic-martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Camus, E.; Wanderka, N.; Welzel, S.; Wollenberger, H. [Hahn-Meitner-Institut Berlin GmbH (Germany); Materna-Morris, E. [Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe (Germany)

    1998-07-15

    Ferritic-martensitic steels containing 8-12 at.% chromium are considered as structural materials for spallation sources and fusion reactors. Materials will be subjected to intense damage rates, e.g. 50-100 dpa per year at full power. Therefore, the behavior under irradiation of these steels must be investigated. Our earlier dual-beam irradiation results on the DIN 1.4914 steel showed a decomposition into chromium-enriched and chromium-depleted regions. The mean concentration of the chromium-depleted regions was found to be 5.19{+-}0.32 at.% after irradiation at 500 C to a fluence of 50 dpa, as measured by atom probe field-ion microscopy. The chromium distribution in the matrix of the DIN 1.4914 steel after thermal ageing at temperatures between 400 and 600 C has been investigated for times up to 17000 h. The carbides were characterized by means of transmission electron microscopy and extraction replicas. The concentrations of the constituents of the matrix were measured by means of atom probe. The mean chromium concentrations in the matrix are found to be 8.66{+-}0.32, 4.5{+-}0.3, and 7.2{+-}0.4 at.%, after ageing at 400, 500, and 600 C, respectively. The matrix contains virtually no carbon. The results are discussed in terms of phase decomposition and species segregation. (orig.) 15 refs.

  9. Effect of 16.3 dpa neutron irradiation on fatigue lifetime of the RAFM steel EUROFER97

    Energy Technology Data Exchange (ETDEWEB)

    Materna-Morris, E., E-mail: edeltraud.materna-morris@kit.edu [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Moeslang, A.; Rolli, R.; Schneider, H.-C. [KIT Karlsruhe Institute of Technology, Campus Nord, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2011-10-15

    Low cycle fatigue specimens of the reduced-activation martensitic/ferritic steel EUROFER97 were neutron irradiated at 250 deg. C up to an accumulated dose of 16.3 dpa. After irradiation, the specimens were push-pull fatigue tested under strain-controlled conditions at 250 deg. C to determine the impact of irradiation on lifetime, fracture behavior, and microstructure. The typical cyclic softening of martensitic/ferritic steels was observed. Furthermore, a considerable increase of lifetime after irradiation and subsequent cycling at lower strain amplitudes was remarkable. This behavior was attributed to the homogeneous distribution of stable irradiation-induced dislocation loops and small precipitates acting as barriers for the cyclic motion of dislocations, thereby influencing substantially crack initiation and crack network formation. While in the un-irradiated material push-pull fatigue sweeps the dislocations to the boundaries, a significant fraction of dislocations was fixed at irradiation-induced defects after irradiation and fatigue testing.

  10. Fracture behaviors of neutron-irradiated ferritic steels studied by the instrumented charpy impact test

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1989-12-01

    The instrumented Charpy impact test for quarter-size specimens was developed and applied to study fracture behavior of ferritic steels and a ferritic-martensitic steel (JFMS) before and after neutron irradiation. The load-deflection curves obtained for U- and V-notched specimens showed typical characteristics of fracture properties of these steels. The temperature dependence of the fracture energy ( Ef) and the failure deflection ( Df) clearly indicates ductile-brittle transition and the DBTT can be determined from the Ef and Df versus temperature curves. The V-notched specimens showed sharper transition at higher temperatures for the JFMS than the U-notched ones, where the former were sensitive to brittle fracture and the latter well demonstrated the behavior of crack propagation. For the ferritic steels the DBTTs showed low values at compositions containing approximate 8-10% Cr and the increase of the DBTT (Δ DBTT) due to irradiation also showed a similar tendency. The Δ DBTT appeared to be relatively larger for the JFMS than the ferritic steels.

  11. 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.

  12. Positron study of steel NF 709 after irradiation and thermal strain

    Science.gov (United States)

    Veternikova, J.; Degmova, J.; Simko, F.; Pekarcikova, M.; Sojak, S.; Slugen, V.

    2015-12-01

    New improved austenitic steel NF 709 was studied in term of thermal and radiation stability in consideration of its application as structural material for the newest generation of nuclear reactors - Generation IV. Samples of steel NF 709 were exposed to two strains: annealing at 1000 °C in argon atmosphere and simulated irradiation performed by helium ion implantation. Changes of the microstructure after the experimental strains were observed by positron annihilation techniques. The microstructure after both treatments indicated growing of vacancy defects; although these changes were small or in the range of error bar. Thus, material NF 709 can be considered as well resistant to these applied strains.

  13. Structural changes in irradiated steels. An example of industrial studies

    Energy Technology Data Exchange (ETDEWEB)

    May, R. [Institut Max von Laue - Paul Langevin (ILL), 38 - Grenoble (France); Miloudi, S. [Electricite de France (France)

    1997-04-01

    The security of nuclear reactors depends on knowledge about the degradation of some of their components by neutron irradiation. It is anticipated in the design of power reactors so as to ensure their safe operation. The long-time behaviour of reactor components is investigated. These studies serve to verify that the predictions concerning mechanical properties of reactors hold, but also to better understand the processes leading to degradation in order to find means for safely increasing the lifetimes of power plants. (author).

  14. Fluence dependence of defect evolution in austenitic stainless steels during fission neutron irradiation

    Science.gov (United States)

    Watanabe, H.; Muroga, T.; Yoshida, N.

    To understand microstructural evolution during fission neutron irradiation, a pure Fe-Cr-Ni ternary alloy, phosphorus-containing model austenitic stainless steels and SUS316 were irradiated in a Japanese Material Testing Reactor (JMTR) at 493 and 613 K. At 493 K, the density of defect cluster increased with the irradiation dose, but there was no significant change in loop density and loop size among all the materials. At 613 K, on the other hand, interstitial type dislocation loops and phosphides were formed in pure ternary and phosphorus-containing alloys, respectively, by an early stage of irradiation. These results suggest that the defect cluster formation at 493 and 613 K is mainly controlled by the cascade damage and long-range migration of free point defects, respectively.

  15. ATR-A1 irradiation experiment on vanadium alloys and low activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tasi, H.; Strain, R.V.; Gomes, I.; Hins, A.G.; Smith, D.L.

    1996-04-01

    To study the mechanical properties of vanadium alloys under neutron irradiation at low temperatures, an experiment was designed and constructed for irradiation in the Advanced Test Reactor (ATR). The experiment contained Charpy, tensile, compact tension, TEM, and creep specimens of vanadium alloys. It also contained limited low-activation ferritic steel specimens as part of the collaborative agreement with Monbusho of Japan. The design irradiation temperatures for the vanadium alloy specimens in the experiment are {approx}200 and 300{degrees}C, achieved with passive gap-gap sizing and fill gas blending. To mitigate vanadium-to-chromium transmutation from the thermal neutron flux, the test specimens are contained inside gadolinium flux filters. All specimens are lithium-bonded. The irradiation started in Cycle 108A (December 3, 1995) and is expected to have a duration of three ATR cycles and a peak influence of 4.4 dpa.

  16. The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part II. Fatigue crack growth rate

    Science.gov (United States)

    Margolin, B.; Minkin, A.; Smirnov, V.; Sorokin, A.; Shvetsova, V.; Potapova, V.

    2016-11-01

    The experimental data on the fatigue crack growth rate (FCGR) have been obtained for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various radiation swelling. The performed study of the fracture mechanisms for cracked specimens under cyclic loading has explained why radiation swelling affects weakly FCGR unlike its effect on fracture toughness. Mechanical modeling of fatigue crack growth has been carried out and the dependencies for prediction of FCGR in irradiated austenitic steel with and with no swelling are proposed and verified with the obtained experimental results. As input data for these dependencies, FCGR for unirradiated steel and the tensile mechanical properties for unirradiated and irradiated steels are used.

  17. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Gomes, I.; Chung, H.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-04-01

    The ATR-A1 irradiation experiment in the Advanced Test Reactor (ATR) was a collaborative U.S./Japan effort to study at low temperatures the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation was completed on May 5, 1996, as planned, after achieving an estimated neutron damage of 4.7 dpa in vanadium. The capsule has since been kept in the ATR water canal for the required radioactivity cool-down. Planning is underway for disassembly of the capsule and test specimen retrieval.

  18. Influence of irradiation on the ductile fracture of a reactor pressure vessel steel

    Science.gov (United States)

    Haušild, Petr; Kytka, Miloš; Karlík, Miroslav; Pešek, Pavel

    2005-05-01

    The mechanical properties of 15Ch2MFA steel were characterised by tensile and instrumented Charpy tests. The fracture surfaces of Charpy specimens broken in the ductile-to-brittle transition temperature range contain a certain proportion of ductile fracture correlated to fracture energy. Measured ductile crack lengths show the same dependence on fracture deflection and/or fracture energy for irradiated and non-irradiated specimens. The decrease of upper shelf energy with increasing neutron fluence could be explained by an increasing amount of shear fracture.

  19. Fracture behavior of neutron-irradiated high-manganese austenitic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Narui, M.; Kayano, H.

    1991-03-01

    The instrumented Charpy impact test was applied to study the fracture behavior of high-manganese austenitic steels before and after neutron irradiations. Quarter-size specimens of a commercial high-manganese steel (18% Mn-5% Ni-16% Cr), three reference steels (21% Mn-1% Ni-9% Cr, 20% Mn-1% Ni-11% Cr, 15% Mn-1% Ni-13% Cr) and two model steels (17% Mn-4.5% Si-6.5% Cr, 22% Mn-4.5% Si-6.5% Cr-0.2% N) were used for the impact tests at temperatures between 77 and 523 K. The load-deflection curves showed typical features corresponding to characteristics of the fracture properties. The temperature dependences of fracture energy and failure deflection obtained from the curves clearly demonstrate only small effects up to 2 × 10 23 n/m 2 ( E > 0.1 MeV) and brittleness at room temperature in 17% Mn-Si-Cr steel at 1.6 × 10 25 n/m 2 ( E > 0.1 MeV), while ductility still remains in 22%Mn-Si-Cr steel.

  20. Polarised SANS study of microstructural evolution under neutron irradiation in a martensitic steel for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coppola, R.; Dewhurst, C.D.; Lindau, R.; May, R.P.; Moeslang, A.; Valli, M

    2004-03-01

    This work presents the results of polarised small-angle neutron scattering (SANS) measurements of modified martensitic steel DIN1.4914, originally developed for application in future fusion reactors (MANET steel). SANS measurements were made using the D22 instrument at the ILL Grenoble using an ad hoc polarised beam set-up. The investigated MANET samples were neutron irradiated and subsequently post-irradiation tempered to reproduce as much as possible the expected service conditions. The results, based on the analysis of the nuclear-magnetic interference, are discussed taking into account both the occurrence of Cr redistribution phenomena with correlated changes in the composition of the precipitate phases, and the growth of non-magnetic defects (He-bubbles or microvoids)

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

    Science.gov (United States)

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

    2014-12-01

    A novel method for grafting of 2-hydroxyethyl methacrylate (HEMA) onto the surface of stainless steel (SS) was explored by using 60Co-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.

  2. Microstructural stability of a self-ion irradiated lanthana-bearing nanostructured ferritic steel

    Energy Technology Data Exchange (ETDEWEB)

    Pasebani, Somayeh [Univ. of Idaho, Moscow, ID (United States). Dept. of Chemical and Materials Engineering; Center for Advanced Energy Studies, Idaho Falls, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States). Dept. of Chemical and Materials Engineering; Center for Advanced Energy Studies, Idaho Falls, ID (United States); Burns, Jatuporn [Center for Advanced Energy Studies, Idaho Falls, ID (United States); Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Alsagabi, Sultan [Univ. of Idaho, Moscow, ID (United States). Dept. of Chemical and Materials Engineering; King Abdulaziz City for Science and Technology, Riyadh (Saudi Arabia). Atomic Energy Research Inst.; Butt, Darryl P. [Center for Advanced Energy Studies, Idaho Falls, ID (United States); Boise State Univ., ID (United States). Dept. of Materials Science and Engineering; Cole, James I. [Center for Advanced Energy Studies, Idaho Falls, ID (United States); Idaho National Lab. (INL), Idaho Falls, ID (United States); Price, Lloyd M. [Texas A & M Univ., College Station, TX (United States). Dept. of Nuclear Engineering; Shao, Lin [Texas A & M Univ., College Station, TX (United States). Dept. of Nuclear Engineering

    2015-07-01

    Thermally stable nanofeatures with high number density are expected to impart excellent high temperature strength and irradiation stability in nanostructured ferritic steels (NFSs) which have potential applications in advanced nuclear reactors. A lanthana-bearing NFS (14LMT) developed via mechanical alloying and spark plasma sintering was used in this study. The sintered samples were irradiated by Fe2+ ions to 10, 50 and 100 dpa at 30 °C and 500 °C. Microstructural and mechanical characteristics of the irradiated samples were studied using different microscopy techniques and nanoindentation, respectively. Overall morphology and number density of the nanofeatures remained unchanged after irradiation. Average radius of nanofeatures in the irradiated sample (100 dpa at 500 °C) was slightly reduced. A notable level of irradiation hardening and enhanced dislocation activity occurred after ion irradiation except at 30 °C and ≥50 dpa. Other microstructural features like grain boundaries and high density of dislocations also provided defect sinks to assist in defect removal.

  3. 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.

  4. The evolution of mechanical property change in irradiated austenitic stainless steels

    Science.gov (United States)

    Lucas, G. E.

    1993-11-01

    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.

  5. Effect of two steel plate's interface on heat transfer under laser beam irradiation

    CERN Document Server

    Zhao Jian Heng; Zhang Shi Wen; Gui Yuan Zhen; Wang Chun Yan; Tang Xiao Song; Zhang Da Yong

    2002-01-01

    It is supposed that there is a gap in the interface of two contacting steel plates due to thermal deformation under laser beam irradiation, and this gap will affect heat transfer in this interface obviously. This supposition is testified by experiments and simulation. This work is helpful to the study of the destruction mechanism under high power laser loading, and provides an effective way for anti-laser research

  6. High resolution grain boundary analysis of neutron irradiated stainless steel using FEG-TEM

    Energy Technology Data Exchange (ETDEWEB)

    Kodama, Mitsuhiro; Ishiyama, Yoshihide; Yokota, Norikatsu [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan)

    1999-09-01

    High-resolution grain boundary analyses of irradiated SUS304 stainless steel using a field emission gun equipped transmission electron microscope were carried out in order to detect radiation-induced grain boundary segregation. The effect of probe size on the measured compositional profiles was studied. The depletion of chromium and enrichment of nickel, phosphorus and silicon were detected at a grain boundary. The measured compositional profiles were affected by the probe size which impeded their interpretation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Chopra, O. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Gruber, Eugene E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shack, William J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2010-06-01

    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

  8. Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Marquis, Emmanuelle [Univ. of Michigan, Ann Arbor, MI (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-03-28

    Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and

  9. Infrared nanosecond pulsed laser irradiation of stainless steel: micro iron-oxide zones generation.

    Science.gov (United States)

    Ortiz-Morales, M; Frausto-Reyes, C; Soto-Bernal, J J; Acosta-Ortiz, S E; Gonzalez-Mota, R; Rosales-Candelas, I

    2014-07-15

    Nanosecond-pulsed, infrared (1064 nm) laser irradiation was used to create periodic metal oxide coatings on the surface of two samples of commercial stainless steel at ambient conditions. A pattern of four different metal oxide zones was created using a galvanometer scanning head and a focused laser beam over each sample. This pattern is related to traverse direction of the laser beam scanning. Energy-dispersive X-ray spectroscopy (EDS) was used to find the elemental composition and Raman spectroscopy to characterize each oxide zone. Pulsed laser irradiation modified the composition of the stainless steel samples, affecting the concentration of the main components within each heat affected zone. The Raman spectra of the generated oxides have different intensity profiles, which suggest different oxide phases such as magnetite and maghemite. In addition, these oxides are not sensible to the laser power of the Raman system, as are the iron oxide powders reported in the literature. These experiments show that it is possible to generate periodic patterns of various iron oxide zones by laser irradiation, of stainless steel at ambient conditions, and that Raman spectroscopy is a useful punctual technique for the analysis and inspection of small oxide areas.

  10. Spherical nanoindentation of proton irradiated 304 stainless steel: A comparison of small scale mechanical test techniques for measuring irradiation hardening

    Science.gov (United States)

    Weaver, Jordan S.; Pathak, Siddhartha; Reichardt, Ashley; Vo, Hi T.; Maloy, Stuart A.; Hosemann, Peter; Mara, Nathan A.

    2017-09-01

    Experimentally quantifying the mechanical effects of radiation damage in reactor materials is necessary for the development and qualification of new materials for improved performance and safety. This can be achieved in a high-throughput fashion through a combination of ion beam irradiation and small scale mechanical testing in contrast to the high cost and laborious nature of bulk testing of reactor irradiated samples. The current work focuses on using spherical nanoindentation stress-strain curves on unirradiated and proton irradiated (10 dpa at 360 °C) 304 stainless steel to quantify the mechanical effects of radiation damage. Spherical nanoindentation stress-strain measurements show a radiation-induced increase in indentation yield strength from 1.36 GPa to 2.72 GPa and a radiation-induced increase in indentation work hardening rate of 10 GPa-30 GPa. These measurements are critically compared against Berkovich nanohardness, micropillar compression, and micro-tension measurements on the same material and similar grain orientations. The ratio of irradiated to unirradiated yield strength increases by a similar factor of 2 when measured via spherical nanoindentation or Berkovich nanohardness testing. A comparison of spherical indentation stress-strain curves to uniaxial (micropillar and micro-tension) stress-strain curves was achieved using a simple scaling relationship which shows good agreement for the unirradiated condition and poor agreement in post-yield behavior for the irradiated condition. The disagreement between spherical nanoindentation and uniaxial stress-strain curves is likely due to the plastic instability that occurs during uniaxial tests but is absent during spherical nanoindentation tests.

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

    Science.gov (United States)

    Chimi, Yasuhiro; Kitsunai, Yuji; Kasahara, Shigeki; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-07-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Roy, R.B.; Solly, B.

    1966-12-15

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

  13. Influence of post irradiation annealing on the mechanical properties and defect structure of AISI 304 steel

    Energy Technology Data Exchange (ETDEWEB)

    Van Renterghem, W.; Van Dyck, S. [SCK-CEN, Mol (Belgium); Al Mazouzi, A. [EDF, Site les renardieres, Moret-sur-Loing (France)

    2011-07-01

    The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24 dpa in the decommissioned Chooz A reactor. The material has been investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as precipitates, most probably (Ni3Si), are present. Martensite phases with a bcc crystal structure were found near the grain boundaries. Annealing at 400 C had almost no effect on the radiation induced defects, but annealing at 500 C resulted in the immediate un-faulting of the Frank loops. As to the mechanical properties, annealing at 400 C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests (SSRT), under pressurised water reactor (PWR) condition, was significantly reduced. Annealing at 500 C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties obtained from SSRT under PWR chemical conditions. (authors)

  14. Influence of post irradiation annealing on the mechanical properties and defect structure of AISI 304 steel

    Energy Technology Data Exchange (ETDEWEB)

    Van Renterghem, W., E-mail: wvrenter@sckcen.be [SCK-CEN, Nuclear Materials Science, Boeretang 200, 2400 Mol (Belgium); Al Mazouzi, A.; Van Dyck, S. [SCK-CEN, Nuclear Materials Science, Boeretang 200, 2400 Mol (Belgium)

    2011-06-15

    The effect of post irradiation annealing on the mechanical properties and the radiation induced defect structure was investigated on stainless steel, of type AISI 304, that was irradiated up to 24 dpa in the decommissioned Chooz A reactor. The material was investigated both in the as-irradiated state as well as after post irradiation annealing. In the as-irradiated specimen the typical radiation induced defects were found as well as {gamma}'-precipitates (Ni{sub 3}Si). Annealing at 400 deg. C had almost no effect on the radiation induced defects, but annealing at 500 deg. C resulted in the immediate unfaulting of the Frank loops. As to the mechanical properties, annealing at 400 deg. C did not strongly affect the yield strength and the ductility of the material, although the fraction of intergranular fracture during slow strain rate tensile tests under pressurised water reactor conditions, was significantly reduced. Annealing at 500 deg. C did reduce the yield strength and restored substantially the ductility and the strain hardening capability of the material. The microstructure investigated by transmission electron microscopy correlates to the mechanical test results. It was found that the observed defect changes after post irradiation annealing provide a reasonable explanation for the observed changes of the mechanical properties.

  15. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Neustroev, V. S.; Ivanov, L. I.; Djomina, E. V.; Platov, Yu. M.

    1991-03-01

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800°C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement.

  16. Irradiation accelerated corrosion of 316L stainless steel in simulated primary water

    Science.gov (United States)

    Raiman, Stephen S.

    The objective of this work is to understand the effects of irradiation on the corrosion of 316L stainless steel in simulated primary water. 316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to simulated PWR primary water to study the effects of radiation on corrosion. A 3.2 MeV proton beam was transmitted through a 37 microm thick sample that served as a "window" into a corrosion cell containing flowing 320° C water with 3 wppm H2. This design permitted radiolysis and displacement damage to occur on the sample surface in contact with the simulated primary water environment. Samples were irradiated for 4, 12, 24, and 72 hrs at dose rates between 400 and 4000 kGy/s, corresponding to damage rates of 7x10-7 to 7x10-6 dpa/s respectively. The structure and composition of the oxide films were characterized using Raman spectroscopy, STEM, and SEM. Sample areas exposed to direct proton irradiation had inner oxide films that were thinner, more porous, and were deficient in chromium when compared to unirradiated oxides. Outer oxides on irradiated samples exhibited a smaller particle size, and had a significant amount of hematite, which was not found on unirradiated samples. The presence of hematite on irradiated samples indicates an increase in electrochemical potential due to irradiation. Dissolution of chromium-rich spinels due to the elevated potential is identified as a likely mechanism behind the loss of inner oxide chromium. It is suggested that the loss of inner-oxide chromium leads to a less protective inner oxide, and a higher rate of oxide dissolution. Sample areas that were not irradiated, but were exposed to the flow of radiolyzed water, exhibited most of the same phenomena found on irradiated areas including loss of Cr and thinner more porous oxides, indicating that water radiolysis is the primary mechanism. When a sample with a pre-formed oxide was irradiated in the same conditions, the region exposed to radiolyzed

  17. Irradiation and inhomogeneity effects on ductility and toughness of (ODS)-7 -13Cr steels

    Energy Technology Data Exchange (ETDEWEB)

    Preininger, D. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany)

    2007-07-01

    Full text of publication follows: The superimposed effect of irradiation defect and structural inhomogeneity formation on tensile ductility and dynamic toughness of ferritic-martensitic 7-13CrW(Mo)VTa(Nb) and oxide dispersion-strengthened (ODS)-7-13CrWVTa(Ti)- RAFM steels has been examined by work hardening and local stress/strain-induced ductile fracture models. Structural inhomogeneities which strongly promoting plastic instability and localized flow might be formed by the applied fabrication process, high dose irradiation and additionally further during deformation by enhanced local dislocation generation around fine particles or due to slip band formation with localized heating at high impact strain rates {epsilon}'. The work hardening model takes into account superimposed dislocation multiplication from stored dislocations, dispersions and also grain boundaries as well as annihilation by cross-slip. Analytical relations have been deduced from the model describing uniform ductility and ductile upper shelf energy (USE) observed from Charpy-impact testes. Especially, the influence of different irradiation defects like atomic clusters, dislocation loops and coherent chromium-rich {alpha}'- precipitates have been considered together with effects from strain rate as well as irradiation (TI) and test temperature TT. Strengthening by clusters and more pronounced by dislocation loops formed at higher TI>250 deg. C reduces uniform ductility and also distinctly stronger dynamic toughness USE. A superimposed hardening by the {alpha}'- formation in higher Cr containing 9-13Cr steels strongly reduces toughness assisted by a combined grain-boundary embrittlement with reduction of the ductile fracture stress. But that improves work hardening and uniform ductility as observed particularly due to nano-scale Y{sub 2}O{sub 3}- dispersions in ODS-RAFM steels. For ODS- steels additionally the strength-induced reduction of toughness is diminished by a combined

  18. Crack growth tests on a ferritic reactor pressure vessel steel under the simultaneous influence of simulated BWR coolant and irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, H. [VGB PowerTech e.V., Essen (Germany); Huettner, F. [Hamburgische Electricitaets-Werke AG, Hamburg (Germany); Ilg, U. [EnBW Kraftwerke AG, Philippsburg (Germany); Wachter, O. [E.ON Kernkraft GmbH, Hannover(Germany); Widera, M. [RWE Power AG, Essen (Germany); Brozova, A.; Ernestova, M.; Kysela, J.; Vsolak, R. [Nuclear Research Institute Rez plc (Czech Republic)

    2004-07-01

    Crack growth tests under constant load with initial in-situ cycling were performed on the low alloy reactor pressure vessel (RPV) steel 22 NiMoCr 3 7 (A 508 Cl. 2) with the goal to determine crack growth rates of irradiated and non-irradiated steel under the simultaneous influence of simulated BWR coolant and irradiation. The tests were performed under conditions as near as possible to operational conditions in a commercial BWR reactor. The research results are summarized and are compared with international data. (orig.)

  19. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Shibayama, T. [Univ. of Hokkaido, Oarai, Ibaraki (Japan). Inst. for Materials Research

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  20. 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

  1. Comparison of four NDT methods for indication of reactor steel degradation by high fluences of neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Tomáš, I., E-mail: tomas@fzu.cz [Institute of Physics ASCR, Na Slovance 2, Prague 18221 (Czech Republic); Vértesy, G. [Research Centre for Natural Sciences, Institute of Technical Physics and Materials Science, Konkoly Thege Miklós út 29-33, H-1121 Budapest (Hungary); Pirfo Barroso, S. [KFKI Atomic Energy Research Institute, Konkoly Thege Miklós út 29-33, H-1121 Budapest (Hungary); The Open University, Walton Hall, MK92BS Milton Keynes (United Kingdom); Kobayashi, S. [Department of Materials Science and Engineering, Faculty of Engineering, Iwate University, Morioka 020-8551 (Japan)

    2013-12-15

    Highlights: • Results of 4 NDT methods on highly irradiated steel are normalized and compared. • Two of the methods (MAT and HV) correlate well with DBTT. • Magnetic Adaptive Testing gives the most sensitive and the best correlated results. • Measurements and sample shapes for an NDT surveillance program are suggested. - Abstract: Results of three magnetic nondestructive methods, Magnetic Barkhausen Emission (MBE), magnetic minor loops Power Scaling Laws (PSL) and Magnetic Adaptive Testing (MAT), and of one reference mechanical measurement, Vickers Hardness (HV), applied on the same series of neutron heavily irradiated nuclear reactor pressure vessel steel materials, were normalized and presented here for the purpose of their straightforward quantitative mutual comparison. It is uncommon to carry out different round-robin testing on irradiated materials, and if not answering all open questions, the comparison alone justifies this paper. The assessment methods were all based on ferromagnetism, although each of them used a different aspect of it. The presented comparison yielded a justified recommendation of the most reliable nondestructive method for indication of the reactor steel irradiation hardening and embrittlement effects. The A533 type B Class 1 steel (JRQ), and the base (15Kh2MFA) and welding (10KhMFT) steels for the WWER 440-type Russian reactors were used for the investigations. The samples were irradiated by high-energy neutrons (>1 MeV) with up to 11.9 × 10{sup 19} n/cm{sup 2} fluences. From all the applied measurements, the results of MAT produced the most satisfactory correlation with independently measured ductile-brittle-transition temperature (DBTT) values of the steel. The other two magnetic methods showed a weaker correlation with DBTT, but some other aspects and information could be assessed by them. As MAT and MBE were sensitive to uncontrolled fluctuation of surface quality of the steel, contact-less ways of testing and more

  2. Accelerated corrosion and oxide dissolution in 316L stainless steel irradiated in situ in high temperature water

    Science.gov (United States)

    Raiman, Stephen S.; Was, Gary S.

    2017-09-01

    316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to high temperature water with added hydrogen (320 °C, 3 wppm H2, neutral pH) to study the effect of radiation on stainless steel corrosion. Irradiated samples had thinner and more porous inner oxides with a lower chromium content when compared to unirradiated samples. Observations suggest that depletion of chromium from the inner oxide can be attributed to the dissolution of chromium-rich spinel oxides in irradiated water, leading to an accelerated rate of inner oxide dissolution. Sample areas which were not irradiated, but were exposed to the flow of irradiated water were also found to be porous and deficient in chromium, indicating that these phenomena can be attributed primarily to water radiolysis. A new empirical equation for oxide growth and dissolution is used to describe the observed changes in oxide thickness under irradiation. An experiment in which a stainless steel sample was exposed to high temperature water (320 °C, 3 wppm H2, neutral pH) without irradiation, and then exposed for a second time with irradiation was conducted to observe the effect of irradiation on a pre-formed protective film. After the irradiated exposure, the sample exhibited chromium loss in regions which were directly irradiated, but not on regions exposed only to irradiated water, suggesting that a pre-formed protective oxide may be effective in preventing chromium loss due to irradiated water. Additionally, this observation suggests that enhanced kinetics under irradiation may have accelerated dissolution of chromium from the inner oxide.

  3. Further Charpy impact test results of low activation ferritic alloys, irradiated at 430{degrees}C to 67 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-04-01

    Miniature CVN specimens of four ferritic alloys, GA3X, F82H, GA4X and HT9, have been impact tested following irradiation at 430{degrees}C to 67 dpa. Comparison of the results with those of the previously tested lower dose irradiation condition indicates that the GA3X and F82H alloys, two primary candidate low activation alloys, exhibit virtually identical behavior following irradiation at 430{degrees}C to {approximately}67 dpa and at 370{degrees}C to {approximately}15 dpa. Very little shift is observed in either DBTT or USE relative to the unirradiated condition. The shifts in DBTT and USE observed in both GA4X and HT9 were smaller after irradiation at 430{degrees}C to {approximately}67 dpa than after irradiation at 370{degrees}C to {approximately}15 dpa.

  4. Interactions between dislocations and irradiation-induced defects in light water reactor pressure vessel steels

    Science.gov (United States)

    Jumel, Stéphanie; Van Duysen, Jean-Claude; Ruste, Jacky; Domain, Christophe

    2005-11-01

    The REVE project (REactor for Virtual Experiments) is an international effort aimed at developing tools to simulate irradiation effects in light water reactors materials. In the framework of this project, a European team developed a first tool, called RPV-1 designed for reactor pressure vessel steels. This article is the third of a series dedicated to the presentation of the codes and models used to build RPV-1. It describes the simplified approach adopted to simulate the irradiation-induced hardening. This approach relies on a characterization of the interactions between a screw dislocation and irradiation-induced defects from molecular dynamics simulations. The pinning forces exerted by the defects on the dislocation were estimated from the obtained results and some hypotheses. In RPV-1, these forces are used as input parameters of a Foreman and Makin-type code, called DUPAIR, to simulate the irradiation-induced hardening at 20 °C. The relevance of the proposed approach was validated by the comparison with experimental results. However, this work has to be considered as an initial step to facilitate the development of a first tool to simulate irradiation effects. It can be improved by many ways (e.g. by use of dislocation dynamics code).

  5. 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; Zhang Yichuan

    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)

  6. Atomistic Analysis Of Radiation-Induced Segregation In Ion-Irradiated Stainless Steel 316

    Directory of Open Access Journals (Sweden)

    Lee G.-G.

    2015-06-01

    Full Text Available Stainless steel (SS is a well-known material for the internal parts of nuclear power plants. It is known that these alloys exhibit radiation-induced segregation (RIS at point defect sinks at moderate temperature, while in service. The RIS behavior of SS can be a potential problem by increasing the susceptibility to irradiation-assisted stress corrosion cracking. In this work, the RIS behavior of solute atoms at sinks in SS 316 irradiated with Fe4+ ions were characterized by atom probe tomography (APT. There were torus-shaped defects along with a depletion of Cr and enrichment of Ni and Si. These clusters are believed to be dislocation loops resulting from irradiation. The segregation of solutes was also observed for various defect shapes. These observations are consistent with other APT results from the literature. The composition of the clusters was analyzed quantitatively almost at the atomic scale. Despite the limitations of the experiments, the APT analysis was well suited for discovering the structure of irradiation defects and performing a quantitative analysis of RIS in irradiated specimens.

  7. 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.

  8. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Science.gov (United States)

    Chiapetto, M.; Messina, L.; Becquart, C. S.; Olsson, P.; Malerba, L.

    2017-02-01

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a "grey-alloy" approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  9. Application of the Master Curve approach for the irradiation embrittlement evaluation of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Viehrig, H.W.; Boehmert, J. [Forschungszentrum Rossendorf e.V., Inst. fuer Sicherheitsforschung, Dresden (Germany)

    2003-09-01

    The master curve (MC) approach and the associated reference temperature, T{sub 0}, as defined in the test standard ASTM E1921, is rapidly moving from the research laboratory to application in integrity assessment of components and structures. T{sub 0} is the index temperature for the universal MC, which considers the toughness behaviour of a specific material. ''The Structural Integrity Assessment Procedures for European Industry'' (SINTAP) contain a MC extension for analysing the fracture behaviour of inhomogeneous ferritic steels. This paper presents the application of the MC approach to the T{sub 0} determination of different types of Russian WWER-type reactor pressure vessel (RPV) steels. In addition the SINTAP-MC approach was applied to determine an alternative reference temperature, T{sub R}. The influence of different microstructures and compositions within one type of RPV steel and the effect of irradiation with fast neutrons on T{sub 0} are experimentally evaluated. In general the MC based T{sub 0} is about 72 K below the Charpy V-notch transition temperature related to an impact energy of 48 J. The paper demonstrates the application of MC based T{sub 0} and T{sub R} as an alternative reference temperature for neutron embrittled RPV steels used in the RPV integrity assessment. (orig.)

  10. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Shack, W. J.; Energy Technology

    2006-01-31

    This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to

  11. Irradiation-assisted stress corrosion cracking behavior of austenitic stainless steels applicable to LWR core internals.

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Shack, W. J.; Energy Technology

    2006-01-31

    This report summarizes work performed at Argonne National Laboratory on irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels that were irradiated in the Halden reactor in simulation of irradiation-induced degradation of boiling water reactor (BWR) core internal components. Slow-strain-rate tensile tests in BWR-like oxidizing water were conducted on 27 austenitic stainless steel alloys that were irradiated at 288 C in helium to 0.4, 1.3, and 3.0 dpa. Fractographic analysis was conducted to determine the fracture surface morphology. Microchemical analysis by Auger electron spectroscopy was performed on BWR neutron absorber tubes to characterize grain-boundary segregation of important elements under BWR conditions. At 0.4 and 1.4 dpa, transgranular fracture was mixed with intergranular fracture. At 3 dpa, transgranular cracking was negligible, and fracture surface was either dominantly intergranular, as in field-cracked core internals, or dominantly ductile or mixed. This behavior indicates that percent intergranular stress corrosion cracking determined at {approx}3 dpa is a good measure of IASCC susceptibility. At {approx}1.4 dpa, a beneficial effect of a high concentration of Si (0.8-1.5 wt.%) was observed. At {approx}3 dpa, however, such effect was obscured by a deleterious effect of S. Excellent resistance to IASCC was observed up to {approx}3 dpa for eight heats of Types 304, 316, and 348 steel that contain very low concentrations of S. Susceptibility of Types 304 and 316 steels that contain >0.003 wt.% S increased drastically. This indicates that a sulfur related critical phenomenon plays an important role in IASCC. A sulfur content of <0.002 wt.% is the primary material factor necessary to ensure good resistance to IASCC. However, for Types 304L and 316L steel and their high-purity counterparts, a sulfur content of <0.002 wt.% alone is not a sufficient condition to ensure good resistance to IASCC. This is in distinct contrast to

  12. IAEA international studies on irradiation embrittlement of reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Brumovsky, M. [Nuclear Research Institute Rez plc (Czech Republic); Steele, L.E. [Chief Scientific Investigator of the Programme, Springfield, VA (United States)

    1997-02-01

    In last 25 years, three phases a Co-operative Research Programme on Irradiation Embrittlement of Reactor Pressure Vessel Steels has been organized by the International Atomic Energy Agency. This programme started with eight countries in 1971 and finally 16 countries took part in phase III of the Programme in 1983. Several main efforts were put into preparation of the programme, but the principal task was concentrated on an international comparison of radiation damage characterization by different laboratories for steels of {open_quotes}old{close_quotes} (with high impurity contents) and {open_quotes}advanced{close_quotes} (with low impurity contents) types as well as on development of small scale fracture mechanics procedures applicable to reactor pressure vessel surveillance programmes. This year, a new programme has been opened, concentrated mostly on small scale fracture mechanics testing.

  13. A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

    Science.gov (United States)

    Chopra, O. K.; Rao, A. S.

    2011-02-01

    Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.

  14. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang; Toloczko, Mychailo B.; Saleh, Tarik A.; Maloy, Stuart A.

    2013-01-14

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa pm occurred in room temperature tests when irradiation temperature was below 400 C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa pm was measured when the irradiation temperature was above 430 C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 *C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  15. Surface and internal microstructure damage of He-ion-irradiated CLAM steel studied by cross-sectional transmission electron microscopy

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P.P. [School of Material Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Zhan, Q., E-mail: qzhan@mater.ustb.edu.cn [School of Material Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Fu, Z.Y.; Wei, Y.P. [School of Material Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Wang, Y.M. [Faculty of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Wang, F.M. [School of Metallurgical and Ecological Engineering, University of Science and Technology Beijing, Beijing 100083 (China); Ohnuki, S. [Faculty of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Wan, F.R. [School of Material Science and Engineering, University of Science and Technology Beijing, Beijing 100083 (China)

    2015-11-15

    Good understanding of blistering and embrittlement mechanism depends on good investigation of surface and internal microstructure damage of gas-ion-irradiated materials. Internal and surface microstructure of He{sup +} ion irradiated CLAM steel were examined by cross-sectional transmission electron microscopy combining focused ion beam. Variation of helium bubble density and size distribution versus depth in CLAM steel after high dose helium irradiation at room temperature was investigated. The average size of helium bubble increased within 100–400 nm but decreased near the non-irradiated matrix with the increase of depth, while the density followed a reverse trend. The formation and growth mechanism of helium bubble is different at different irradiation depth. The formation of a zone of large bubbles under the surface is the main reason of surface blistering and flaking. Helium induced irradiation swelling and surface blistering at low temperature were also discussed. - Highlights: • Microstructure of helium irradiated CLAM steel was investigated by FIB and TEM. • The nucleation of helium bubble was controlled by both different mechanisms. • The substructure of surface blisters has been analyzed in detail by XTEM. • Helium induced surface blistering and irradiation swelling have been discussed.

  16. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    2013-01-01

    The present work concerns investigations of damage accumulation during helium implantation of pure iron and the reduced activation ferritic-martensitic steel 'EUROFER 97' at 323K and 623K as well as during neutron irradiation with or without prior helium implantation. The defect microstructure......, in particular the cavities, was characterized using Positron Annihilation Lifetime Spectroscopy (PALS) and Transmission Electron Microscopy (TEM). The PALS investigations reveal a clear difference between the He implantation effects in Fe and EUROFER 97 at both temperatures. For both materials the mean positron...

  17. Change in the properties of Fe-Cr-Ni and Fe-Cr-Mn austenitic steels under mixed and fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Bulanova, T.M. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Golovanov, V.N. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Neustroyev, V.S. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Povstyanko, A.V. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors; Ostrovsky, Z.E. [State Sci. Centre of Russian Federation, Dimitrovgrad (Russian Federation). Res. Inst. of Atomic Reactors

    1996-10-01

    Detailed investigations are performed on mechanical properties, swelling and structure of different types of Fe-Cr-Ni and Fe-Cr-Mn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800 C and the maximum achieved level of damage doses is 60 dpa for Fe-Cr-Mn steel (with 4-5% of Ni) and 30 dpa for steels of the C-12Cr-20Mn-W-T type. Presented are dose dependencies of swelling and mechanical properties of Fe-Cr-Ni and Fe-Cr-Mn steels. It is shown that at temperatures below 530 C the investigated Fe-Cr-Mn steel systems are less susceptible to swelling as compared to Fe-Cr-Ni ones. Fe-Cr-Mn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400 C and much higher embrittlement after irradiation from 350 up to 400 C in the fast spectrum in comparison with Fe-Cr-Ni steels. Higher hardening rate of Fe-Cr-Mn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in Fe-Cr-Mn steels under irradiation are considered taking into account austenite stabilization in the initial state. (orig.).

  18. Change in the properties of FeCrNi and FeCrMn austenitic steels under mixed and fast neutron irradiation

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Golovanov, V. N.; Neustroyev, V. S.; Povstyanko, A. V.; Ostrovsky, Z. E.

    1996-10-01

    Detailed investigations are performed on mechanical properties, swelling and structure of different types of FeCrNi and FeCrMn austenitic stainless steels irradiated in the SM-2 high-flux research reactor and BOR-60 fast reactor. Steel irradiation temperatures are ranging from 100 up to 800°C and the maximum achieved level of damage doses is 60 dpa for FeCrMn steel (with 4-5% of Ni) and 30 dpa for steels of the C12Cr20MnWT type. Presented are dose dependencies of swelling and mechanical properties of FeCrNi and FeCrMn steels. It is shown that at temperatures below 530°C the investigated FeCrMn steel systems are less susceptible to swelling as compared to FeCrNi ones. FeCrMn steels showed a lower value of irradiation embrittlement after irradiation in the mixed spectrum at temperatures from 100 up to 400°C and much higher embrittlement after irradiation from 350 up to 400°C in the fast spectrum in comparison with FeCrNi steels. Higher hardening rate of FeCrMn steels after their irradiation in BOR-60 is attributed to the presence of dislocation loops and defects of high density in the structure. The structural change features in FeCrMn steels under irradiation are considered taking into account austenite stabilization in the initial state.

  19. A study on martensitic and austenitic steels after exposure in mercury at 573 K up to 5000 h

    Energy Technology Data Exchange (ETDEWEB)

    Zalavutdinov, R.Kh. E-mail: rinad@ipc.rssi.ru; Dai, Y.; Gorodetsky, A.E.; Bauer, G.S.; Alimov, V.Kh.; Zakharov, A.P

    2001-07-01

    The chemical composition, structure and morphology of surface layers formed on stressed martensitic (F82H, MANET-II) and austenitic (316L) steel samples after exposure in static mercury with air or Ar inside the containers for 5000 and 2000 h, respectively, at 573 K have been studied by different surface analysis techniques (electron probe microanalysis (EPMA), scanning electron microscopy (SEM), reflected high energy electron diffraction (RHEED), X-ray diffraction (XRD), and secondary ion mass spectrometry (SIMS)). It has been shown that all three steels are oxidized (oxide thickness greatest on F82H and least for 316L) and covered with red HgO single crystals when air is present in the system. The oxidation of the steels and Hg can be suppressed by using Ar in the containers. Cracks have been found only at the notch roots of the 316L samples. There is no evident Hg corrosion observed.

  20. Status of ATR-A1 irradiation experiment on vanadium alloys and low-activation steels

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Strain, R.V.; Gomes, I.; Smith, D.L. [Argonne National Lab., IL (United States); Matsui, H. [Tohoku Univ. (Japan)

    1996-10-01

    The ATR-A1 irradiation experiment was a collaborative U.S./Japan effort to study at low temperature the effects of neutron damage on vanadium alloys. The experiment also contained a limited quantity of low-activation ferritic steel specimens from Japan as part of the collaboration agreement. The irradiation started in the Advanced Test Reactor (ATR) on November 30, 1995, and ended as planned on May 5, 1996. Total exposure was 132.9 effective full power days (EFPDs) and estimated neutron damage in the vanadium was 4.7 dpa. The vehicle has been discharged from the ATR core and is scheduled to be disassembled in the next reporting period.

  1. 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.

  2. 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.

  3. Convoluted dislocation loops induced by helium irradiation in reduced-activation martensitic steel and their impact on mechanical properties

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen' s University, Kingston, ON, Canada K7L 3N6 (Canada); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Wen, Yongming [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory, School of Physics and Technology, Wuhan University, Wuhan 430072 (China)

    2014-06-01

    Helium irradiation induced dislocation loops in reduced-activation martensitic steels were investigated using transmission electron microscopy. The specimens were irradiated with 100 keV helium ions to 0.8 dpa at 350 °C. Unexpectedly, very large dislocation loops were found, significantly larger than that induced by other types of irradiations under the same dose. Moreover, the large loops were convoluted and formed interesting flower-like shape. The large loops were determined as interstitial type. Loops with the Burgers vectors of b=〈100〉 were only observed. Furthermore, irradiation induced hardening caused by these large loops was observed using the nano-indentation technique.

  4. SANS examination of irradiated RPV steel welds during in-situ annealing

    Energy Technology Data Exchange (ETDEWEB)

    Boothby, R.M. [National Nuclear Laboratory, B168 Harwell Campus, Didcot, Oxon. OX11 0QT (United Kingdom); Hyde, J.M., E-mail: jonathan.m.hyde@nnl.co.uk [National Nuclear Laboratory, B168 Harwell Campus, Didcot, Oxon. OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Swan, H. [National Nuclear Laboratory, B168 Harwell Campus, Didcot, Oxon. OX11 0QT (United Kingdom); Parfitt, D.; Wilford, K. [Rolls-Royce plc, P.O. Box 2000, Derby DE21 7XX (United Kingdom); Lindner, P. [Institut Laue-Langevin, 71 avenue des Martyrs, CS20156, 38042 Grenoble CEDEX 9 (France)

    2015-06-15

    An in-situ annealing experiment was performed using SANS measurements to examine the distribution and thermal stability of irradiation-induced solute clusters in RPV steel welds. Samples were sequentially annealed for 30 min at ∼50 °C intervals in the temperature range 295–497 °C. A methodology was developed to correct the observed data to allow for increased thermal diffuse scattering during annealing which enabled analysis of the changes in coherent scattering in isolation. Results for a low-Ni weld irradiated at low temperature showed apparent decreases in the volume fraction of solute clusters during annealing. However the cluster size was unaffected and these results could have arisen from reduced scattering contrast due to compositional changes, rather than cluster dissolution. A similarly irradiated high-Ni weld exhibited cluster coarsening at high annealing temperatures. Samples of both welds irradiated at a higher temperature were relatively unaffected by annealing except at high temperatures where some shrinkage, indicative of cluster dissolution, occurred.

  5. Void Swelling and Microstructure of Austenitic Stainless Steels Irradiated in the BOR - 60 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, Yong [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, Yina [Argonne National Lab. (ANL), Argonne, IL (United States); Allen, T. [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)

    2012-11-01

    As nuclear power plants age and neutron fluence increases, detrimental effects resulting from radiation damage have become an increasingly important issue for the operational safety and structural integrity of core internal components. In this study, irradiated specimens of reactor core internal components were characterized by transmission electron microscopy. The specimens had been irradiated to 5.5-45 dpa in the BOR-60 reactor at a dose rate close to 10-6 dpa/s and temperature of about 320°C. No voids were observed in the austenitic stainless steels and nickel alloys at all doses. Despite the possibility that fine voids below the TEM resolution limit may be present, it was clear that void swelling was insignificant in all examined alloys up to 45 dpa. Irradiated microstructures of the studied alloys were dominated by a high density of Frank loops. The mean size and density of the Frank loops varied from one material to another, but saturated with increasing dose above ~10 dpa. While no irradiation-induced precipitations were present below 24.5 dpa, fine precipitates were evident in several alloys at 45 dpa.

  6. Stability of the strengthening nanoprecipitates in reduced activation ferritic steels under Fe2+ ion irradiation

    Science.gov (United States)

    Tan, L.; Katoh, Y.; Snead, L. L.

    2014-02-01

    The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic-martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ∼49 dpa at 500 °C using Fe2+ is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe2W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.

  7. Helium effects on mechanical properties and microstructure of high fluence ion-irradiated RAFM steel

    Science.gov (United States)

    Ogiwara, H.; Kohyama, A.; Tanigawa, H.; Sakasegawa, H.

    2007-08-01

    Reduced-activation ferritic/martensitic steels, RAFS, are leading candidates for the blanket and first wall of fusion reactors, and effects of displacement damage and helium production on mechanical properties and microstructures are important to these applications. Because it is the most effective way to obtain systematic and accurate information about microstructural response under fusion environment, single-(Fe 3+) and dual-(Fe 3+ + He +) irradiations were performed followed by TEM observation and nano-indentation hardness measurement. Dual-ion irradiation at 420 °C induced finer defect clusters compared to single-ion irradiation. These fine defect clusters caused large differences in the hardness increase between these irradiations. TEM analysis clarified that radiation induced precipitates were MX precipitates (M: Ta, W). Small defects invisible to TEM possibly caused the large increase in hardness, in addition to the hardness increment produced by radiation induced MX. In this work, radiation hardening and microstructural evolution accompanied by the synergistic effects to high fluences are discussed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  9. Self-ordered defect structures in two model F/M steels under in situ ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Kaoumi, D., E-mail: djamelkaoumi@gmail.com; Adamson, J.

    2014-05-01

    Two model F/M steels, 9Cr-model and 12Cr-model, were irradiated with 1 MeV Kr ions in situ in a TEM at temperatures between 20 K and 573 K to doses as high as 15 dpa. During the early stages of irradiation of the two F/M steels, defect clusters were rather uniformly distributed within grains, and a saturation density was quickly reached. However, at higher doses, self-ordering alignments of defect clusters were found in some grains. The regularly ordered arrays of small loops were observed in the two F/M steels along 〈1 1 0〉 directions with spacing about 30–50 nm. Once the aligned structure was created, it was stable under further irradiation. The possible mechanisms for the “self-organization”/“ordering” of the clusters were investigated. This paper describes the process and its temperature dependence, and the possible mechanisms are discussed.

  10. Cracking behavior and microstructure of austenitic stainless steels and alloy 690 irradiated in BOR-60 reactor, phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Chopra, O. K.; Soppet, W. K.; Shack, W. J.; Yang, Y.; Allen, T. R.; Univ. of Wisconsin at Madison

    2010-02-16

    Cracking behavior of stainless steels specimens irradiated in the BOR-60 at about 320 C is studied. The primary objective of this research is to improve the mechanistic understanding of irradiation-assisted stress corrosion cracking (IASCC) of core internal components under conditions relevant to pressurized water reactors. The current report covers several baseline tests in air, a comparison study in high-dissolved-oxygen environment, and TEM characterization of irradiation defect structure. Slow strain rate tensile (SSRT) tests were conducted in air and in high-dissolved-oxygen (DO) water with selected 5- and 10-dpa specimens. The results in high-DO water were compared with those from earlier tests with identical materials irradiated in the Halden reactor to a similar dose. The SSRT tests produced similar results among different materials irradiated in the Halden and BOR-60 reactors. However, the post-irradiation strength for the BOR-60 specimens was consistently lower than that of the corresponding Halden specimens. The elongation of the BOR-60 specimens was also greater than that of their Halden specimens. Intergranular cracking in high-DO water was consistent for most of the tested materials in the Halden and BOR-60 irradiations. Nonetheless, the BOR-60 irradiation was somewhat less effective in stimulating IG fracture among the tested materials. Microstructural characterization was also carried out using transmission electron microscopy on selected BOR-60 specimens irradiated to {approx}25 dpa. No voids were observed in irradiated austenitic stainless steels and cast stainless steels, while a few voids were found in base and grain-boundary-engineered Alloy 690. All the irradiated microstructures were dominated by a high density of Frank loops, which varied in mean size and density for different alloys.

  11. Heavy-Section Steel Irradiation Program: Progress report for April--September 1995. Volume 6, Number 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1996-08-01

    The goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of effects of neutron irradiation on material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and post-irradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness (K{sub Ic}) curve shift in high-copper welds, (3) crack-arrest toughness (K{sub Ia}) curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub Ic} and K{sub Ia} curve shifts in low upper-shelf welds, (6) annealing effects in low upper-shelf welds, (7) irradiation effects in a commercial low upper-shelf weld, (8) microstructural analysis of irradiation effects, (9) in-service aged material evaluations, (10) correlation monitor materials, (11) special technical assistance, (12) JPDR steel examination, (13) technical assistance for JCCCNRS Working Groups 3 and 12, and (14) additional requirements for materials. This report provides an overview of the activities within each of these task from April through September 1995.

  12. True stress–strain curve acquisition for irradiated stainless steel including the range exceeding necking strain

    Energy Technology Data Exchange (ETDEWEB)

    Kamaya, Masayuki, E-mail: kamaya@inss.co.jp [Institute of Nuclear Safety System, Inc., 64 Sata Mihama-cho, Fukui 919-1205 (Japan); Kitsunai, Yuji; Koshiishi, Masato [Nippon Nuclear Fuel Dvelopment Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan)

    2015-10-15

    True stress–strain curves were obtained for irradiated 316L stainless steel by a tensile test and by a curve estimation procedure. In the tensile test, the digital image correlation technique together with iterative finite element analysis was applied in order to identify curves for strain larger than the necking strain. The true stress–strain curves were successfully obtained for the strain of more than 0.4 whereas the necking strain was about 0.2 in the minimum case. The obtained true stress–strain curves were approximated well with the Swift-type equation including the post-necking strain even if the exponential constant n was fixed to 0.5. Then, the true stress–strain curves were estimated by a curve estimation procedure, which was referred to as the K-fit method. Material properties required for the K-fit method were the yield and ultimate strengths or only the yield strength. Some modifications were made for the K-fit method in order to improve estimation accuracy for irradiated stainless steels.

  13. Evolution of structure and properties of VVER-1000 RPV steels under accelerated irradiation up to beyond design fluences

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Maltsev, D.; Frolov, A.; Zabusov, O.; Erak, D.; Zhurko, D.

    2015-01-01

    In this paper comprehensive studies of structure and properties of VVER-1000 RPV steels after the accelerated irradiation to fluences corresponding to extended lifetime up to 60 years or more as well as comparative studies of materials irradiated with different fluxes were carried out. The significant flux effect is confirmed for the weld metal (nickel concentration ⩾1.35%) which is mainly due to development of reversible temper brittleness. The rate of radiation embrittlement of VVER-1000 RPV steels under operation up to 60 years and more (based on the results of accelerated irradiation considering flux effect for weld metal) is expected not to differ significantly from the observed rate under irradiation within surveillance specimens.

  14. Effects of Ti element on the microstructural stability of 9Cr–WVTiN reduced activation martensitic steel under ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Fengfeng [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Jin, Shuoxue; Li, Tiecheng; Chen, Jihong [Key Laboratory of Artificial Micro- and Nano-Structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Yao, Z. [Department of Mechanical and Materials Engineering, Queen’s University, Kingston K7L 3N6, ON (Canada)

    2014-12-15

    Microstructure of 9Cr–WVTiN reduced-activation martensitic steels with two different Ti concentrations irradiated with Fe{sup +}, He{sup +} and H{sup +} at 300 °C was studied with transmission electron microscopy. Small dislocation loops were observed in the irradiated steels. The mean size and number density of dislocation loops decreased with the increase of Ti concentration. The segregation of Cr and Fe in carbides was observed in both irradiated steels, and the enrichment of Cr and depletion of Fe were more severe in the low Ti-concentration 9Cr–WVTiN steel.

  15. Structural Transformations in Austenitic Stainless Steel Induced by Deuterium Implantation: Irradiation at 295 K

    Science.gov (United States)

    Morozov, Oleksandr; Zhurba, Volodymir; Neklyudov, Ivan; Mats, Oleksandr; Progolaieva, Viktoria; Boshko, Valerian

    2016-02-01

    Deuterium thermal desorption spectra were investigated on the samples of austenitic steel 18Cr10NiTi pre-implanted at 295 K with deuterium ions in the dose range from 8 × 1014 to 2.7 × 1018 D/cm2. The kinetics of structural transformation development in the steel layer was traced from deuterium thermodesorption spectra as a function of deuterium concentration. Three characteristic regions with different low rates of deuterium amount desorption as the implantation dose increases were revealed: I—the linear region of low implantation doses (up to 1 × 1017 D/cm2); II—the nonlinear region of medium implantation doses (1 × 1017 to 8 × 1017 D/cm2); III—the linear region of high implantation doses (8 × 1017 to 2.7 × 1018 D/cm2). During the process of deuterium ion irradiation, the coefficient of deuterium retention in steel varies in discrete steps. Each of the discrete regions of deuterium retention coefficient variation corresponds to different implanted-matter states formed during deuterium ion implantation. The low-dose region is characterized by formation of deuterium-vacancy complexes and solid-solution phase state of deuterium in the steel. The total concentration of the accumulated deuterium in this region varies between 2.5 and 3 at.%. The medium-dose region is characterized by the radiation-induced action on the steel in the presence of deuterium with the resulting formation of the energy-stable nanosized crystalline structure of steel, having a developed network of intercrystalline boundaries. The basis for this developed network of intercrystalline boundaries is provided by the amorphous state, which manifests itself in the thermodesorption spectra as a widely temperature-scale extended region of deuterium desorption (structure formation with a varying activation energy). The total concentration of the accumulated deuterium in the region of medium implantation doses makes 7 to 8 at.%. The resulting structure shows stability against the action of

  16. 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.

  17. Microstructural evolution of ferritic-martensitic steels under heavy ion irradiation

    Science.gov (United States)

    Topbasi, Cem

    Ferritic-martensitic steels are primary candidate materials for fuel cladding and internal applications in the Sodium Fast Reactor, as well as first-wall and blanket materials in future fusion concepts because of their favorable mechanical properties and resistance to radiation damage. Since microstructure evolution under irradiation is amongst the key issues for these materials in these applications, developing a fundamental understanding of the irradiation-induced microstructure in these alloys is crucial in modeling and designing new alloys with improved properties. The goal of this project was to investigate the evolution of microstructure of two commercial ferritic-martensitic steels, NF616 and HCM12A, under heavy ion irradiation at a broad temperature range. An in situ heavy ion irradiation technique was used to create irradiation damage in the alloy; while it was being examined in a transmission electron microscope. Electron-transparent samples of NF616 and HCM12A were irradiated in situ at the Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory with 1 MeV Kr ions to ˜10 dpa at temperatures ranging from 20 to 773 K. The microstructure evolution of NF616 and HCM12A was followed in situ by systematically recording micrographs and diffraction patterns as well as capturing videos during irradiation. In these irradiations, there was a period during which no changes are visible in the microstructure. After a threshold dose (˜0.1 dpa between 20 and 573 K, and ˜2.5 dpa at 673 K) black dots started to become visible under the ion beam. These black dots appeared suddenly (from one frame to the next) and are thought to be small defect clusters (2-5 nm in diameter), possibly small dislocation loops with Burgers vectors of either ½ or . The overall density of these defect clusters increased with dose and saturated around 6 dpa. At saturation, a steady-state is reached in which defects are eliminated and created at the same rates so that the

  18. Deformation behavior of reduced activation ferritic steel during tensile test

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, Kiyoyuki [Department of Material Science and Engineering, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibakaki 319-1195 (Japan)]. E-mail: shiba@realab01.tokai.jaeri.go.jp; Hirose, Takanori [Department of Fusion Engineering Research, Japan Atomic Energy Research Institute, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

    2006-02-15

    Deformation behavior of reduced activation martensitic steel F82H during tensile tests were studied. True stress-true strain diagrams were calculated with minimum diameter determined from the specimen profile obtained by laser micro-gauge scanning the diameter along the longitudinal direction during tensile test. Cylindrical specimens of F82H were used for the measurement and test temperatures were room temperature (RT), 300, 400, 500 and 600 deg. C. Tensile tests were carried out with 1 x 10{sup -4} s{sup -1} of strain rate. Other strain rates (1 x 10{sup -3} and 1 x 10{sup -5} s{sup -1}) were applied for the tests at RT. Although uniform elongation of F82H is relatively small at elevated temperature, true stress increases to fracture after necking starts. True stress decreases temporarily after yielding at 600 deg. C, but it increases again to fracture like the specimens tested at lower temperatures. Influence of strain rate to true stress-true strain relationship at room temperature was small, but unstable deformation occurred in narrower area at higher strain rate.

  19. Evolution of the mechanical properties and microstructure of ferritic-martensitic steels irradiated in the BOR-60 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K. E-mail: fae@niiar.ru; Golovanov, V.N.; Bulanova, T.M.; Povstyanko, A.V.; Fedoseev, A.E.; Ostrovsky, Z.E.; Goncharenko, Yu.D

    2002-12-01

    The effect of neutron irradiation on mechanical properties of low-activation ferritic-martensitic (FM) steels 0.1C-9Cr-1W, V, Ta, B and 0.1C-12Cr-2W, V, Ti, B is studied under tension at temperatures of 330-540 deg. C and doses of 50 dpa. Steel 0.1C-13Cr-Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330-340 deg. C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 deg. C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model {epsilon}-bar/{sigma}-bar=B{sub 0}+DS, when B{sub 0} and D have the values typical for steels of FM type.

  20. Evolution of the mechanical properties and microstructure of ferritic-martensitic steels irradiated in the BOR-60 reactor

    Science.gov (United States)

    Shamardin, V. K.; Golovanov, V. N.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.; Ostrovsky, Z. E.; Goncharenko, Yu. D.

    2002-12-01

    The effect of neutron irradiation on mechanical properties of low-activation ferritic-martensitic (FM) steels 0.1C-9Cr-1W, V, Ta, B and 0.1C-12Cr-2W, V, Ti, B is studied under tension at temperatures of 330-540 °C and doses of 50 dpa. Steel 0.1C-13Cr-Mo, V, Nb, B was chosen for comparison. At irradiation temperatures of 330-340 °C, the radiation hardening of steel with 9%Cr achieves saturation at a dose of 10 dpa. In this case as compared to steels with 12%Cr, the fracture surface is characterized as ductile without cleavage traces. At irradiation temperatures higher than 420 °C, there is no difference in the behavior of the materials under investigation. The data on radiation creep obtained by direct measurement and from the profilometry data satisfy a model ɛ¯/ σ¯=B 0+D Ṡ, when B0 and D have the values typical for steels of FM type.

  1. Neutron diffraction analysis of Cr-Ni-Mo-Ti austenitic steel after cold plastic deformation and fast neutrons irradiation

    Science.gov (United States)

    Voronin, V. I.; Valiev, E. Z.; Berger, I. F.; Goschitskii, B. N.; Proskurnina, N. V.; Sagaradze, V. V.; Kataeva, N. F.

    2015-04-01

    A quantitative assessment is presented of the dislocation density and relative fractions of edge and screw dislocations in reactor-steel samples 16Cr-15Ni-3Mo-1Ti subjected to preliminary cold deformation by rolling and subsequent fast neutron irradiation using neutron diffraction analysis. The Williamson-Hall modified method was used for calculations. It is shown that the fast neutron irradiation leads to a decrease in the density of dislocations that appeared after samples deformation. The applicability of neutron diffraction analysis to the examination of dislocation structure of deformed and irradiated materials is shown.

  2. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50-400)°C

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Bukina, Z. V.; Frolov, A. S.; Maltsev, D. A.; Krikun, E. V.; Zhurko, D. A.; Zhuchkov, G. M.

    2017-07-01

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50-400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects - dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

  3. Effect of Grain Size on Void Formation during High-Energy Electron Irradiation of Austenitic Stainless Steel

    DEFF Research Database (Denmark)

    Singh, Bachu Narain

    1974-01-01

    Thin foils of an ‘ experimental ’ austenitic stainless steel, with and without dispersions of aluminium oxide particles, are irradiated with 1 MeV electrons in a High Voltage Electron Microscope at 600°C. Evidence of grain size dependent void nucleation, void concentration, and void volume swelling...

  4. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K.; Bulanova, T.M.; Neustroyev, V.S.; Ostrovsky, Z.E.; Kosenkov, V.M. (V.I. Lenin Research Inst. of Atomic Reactors, Dimitrovgrad (Russia)); Ivanov, L.I.; Djomina, E.V. (A.A. Baikov Inst. of Metallurgy, Academy of Science, Moscow (Russia))

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperture range 400-800deg C (every 100deg C) to 16 dpa dose with 1000 and 3000 appm helium generation, correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100deg C to 21 dpa exhibited significant radiation hardening accompanied by [alpha]-phase formation in the steel structure. (orig.).

  5. Study of Fe-12Cr-20Mn-W-C austenitic steels irradiated in the SM-2 reactor

    Science.gov (United States)

    Shamardin, V. K.; Bulanova, T. M.; Neustroyev, V. S.; Ostrovsky, Z. E.; Kosenkov, V. M.; Ivanov, L. I.; Djomina, E. V.

    1992-09-01

    A comparison has been made between the mechanical properties and swelling of austenitic stainless steels EP-838 (Fe-Cr-Mn) and 316SS (Fe-Cr-Ni) irradiated in the mixed-neutron spectrum of the SM-2 reactor in the temperature range 400-800°C (every 100°C) to 16 dpa dose with 1000 and 3000 appm helium generation correspondingly, determined by nickel content. EP-838 exhibited less susceptibility to void swelling and radiation hardening. Fe-12Cr-20Mn-W-0.1C steel without nickel irradiated at 100°C to 21 dpa exhibited significant radiation hardening accompanied by α-phase formation in the steel structure.

  6. Results of crack-arrest tests on irradiated a 508 class 3 steel

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Milella, P.P.; Pini, M.A.

    1998-02-01

    Ten crack-arrest toughness values for irradiated specimens of A 508 class 3 forging steel have been obtained. The tests were performed according to the American Society for Testing and Materials (ASTM) Standard Test Method for Determining Plane-Strain Crack-Arrest Fracture Toughness, K{sub la} of Ferritic Steels, E 1221-88. None of these values are strictly valid in all five ASTM E 1221-88 validity criteria. However, they are useful when compared to unirradiated crack-arrest specimen toughness values since they show the small (averaging approximately 10{degrees}C) shifts in the mean and lower-bound crack-arrest toughness curves. This confirms that a low copper content in ASTM A 508 class 3 forging material can be expected to result in small shifts of the transition toughness curve. The shifts due to neutron irradiation of the lower bound and mean toughness curves are approximately the same as the Charpy V-notch (CVN) 41-J temperature shift. The nine crack-arrest specimens were irradiated at temperatures varying from 243 to 280{degrees}C, and to a fluence varying from 1.7 to 2.7 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV). The test results were normalized to reference values that correspond to those of CVN specimens irradiated at 284{degrees}C to a fluence of 3.2 x 10{sup 19} neutrons/cm{sup 2} (> 1 MeV) in the same capsule as the crack-arrest specimens. This adjustment resulted in a shift to lower temperatures of all the data, and in particular moved two data points that appeared to lie close to or lower than the American Society of Mechanical Engineers K{sub la} curve to positions that seemed more reasonable with respect to the remaining data. A special fixture was designed, fabricated, and successfully used in the testing. For reasons explained in the text, special blocks to receive the Oak Ridge National Laboratory clip gage were designed, and greater-than-standard crack-mouth opening displacements measured were accounted for. 24 refs., 13 figs., 12 tabs.

  7. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    Science.gov (United States)

    Chernov, I. I.; Kalashnikov, A. N.; Kalin, B. A.; Binyukova, S. Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 × 10 20 m -2 at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content ( NC>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  8. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Chernov, I.I. E-mail: chernov@phm.mephi.ru; Kalashnikov, A.N.; Kalin, B.A.; Binyukova, S.Yu

    2003-12-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe-C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion irradiation up to a fluence of 5 x 10{sup 20} m{sup -2} at the temperature of 920 K. It was shown that helium-ion irradiation at high temperature resulted in formation of bubbles with a greater size and a smaller density in Fe and ferritic-martensitic steels than those in nickel and austenitic steels. Large gaseous bubbles in ferritic component are uniformly distributed in grains body in Fe-C alloys as well as in ferritic-martensitic steels. The bubbles with a higher density and a smaller size than those in ferritic component are formed in martensitic grains of steels and Fe-C alloys with a high carbon content (N{sub C}>0.01 wt%), which leads to a small level of swelling of martensite in comparison with that of ferrite. In addition, the bubbles in martensitic grains have a tendency to ordered distribution.

  9. Effects of low temperature neutron irradiation on deformation behavior of austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, J.E.; Rowcliffe, A.F.; Alexander, D.J.; Grossbeck, M.L. [Oak Ridge National Laboratory, TN (United States); Shiba, K.

    1996-04-01

    An austenitic stainless steel, designated 316LN-IG, has been chosen for the first wall/shield (FW/S) structure for the International Thermonuclear Experimental Reactor (ITER). The proposed operational temperature range for the structure (100 to 250{degree}C) is below the temperature regimes for void swelling (400-600{degree}C) and for helium embrittlement (500-700{degree}C). However, the proposed neutron dose is such that large changes in yield strength, deformation mode, and strain hardening capacity could be encountered which could significantly affect fracture properties. Definition of the irradiation regimes in which this phenomenon occurs is essential to the establishment of design rules to protect against various modes of failure.

  10. In situ micro-tensile testing on proton beam-irradiated stainless steel

    Science.gov (United States)

    Vo, H. T.; Reichardt, A.; Frazer, D.; Bailey, N.; Chou, P.; Hosemann, P.

    2017-09-01

    Small-scale mechanical testing techniques are currently being explored and developed for engineering applications. In particular, micro-tensile testing can add tremendous value, since the entire stress-strain curve, including the strain to failure, can be measured directly. In this work, 304 stainless steel specimens irradiated with 2 MeV protons to 10 dpa (full-cascade setting in the Stopping and Range of Ions in Matter, SRIM, software) at 360 °C was evaluated using micro-tensile testing. It was found that even on the micron scale, the measured strain corresponds well with macroscopic expectations. In addition, a new approach to analyzing sudden slip events is presented.

  11. Composite model of microstructural evolution in austenitic stainless steel under fast neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.; Odette, G.R.

    1986-01-01

    A rate-theory-based model has been developed which includes the simultaneous evolution of the dislocation and cavity components of the microstructure of irradiated austenitic stainless steels. Previous work has generally focused on developing models for void swelling while neglecting the time dependence of the dislocation structure. These models have broadened our understanding of the physical processes that give rise to swelling, e.g., the role of helium and void formation from critically-sized bubbles. That work has also demonstrated some predictive capability by successful calibration to fit the results of fast reactor swelling data. However, considerable uncertainty about the values of key parameters in these models limits their usefulness as predictive tools. Hence the use of such models to extrapolate fission reactor swelling data to fusion reactor conditions is compromised.

  12. Parametric study of irradiation effects on the ductile damage and flow stress behavior in ferritic-martensitic steels

    Science.gov (United States)

    Chakraborty, Pritam; Biner, S. Bulent

    2015-10-01

    Ferritic-martensitic steels are currently being considered as structural materials in fusion and Gen-IV nuclear reactors. These materials are expected to experience high dose radiation, which can increase their ductile to brittle transition temperature and susceptibility to failure during operation. Hence, to estimate the safe operational life of the reactors, precise evaluation of the ductile to brittle transition temperatures of ferritic-martensitic steels is necessary. Owing to the scarcity of irradiated samples, particularly at high dose levels, micro-mechanistic models are being employed to predict the shifts in the ductile to brittle transition temperatures. These models consider the ductile damage evolution, in the form of nucleation, growth and coalescence of voids; and the brittle fracture, in the form of probabilistic cleavage initiation, to estimate the influence of irradiation on the ductile to brittle transition temperature. However, the assessment of irradiation dependent material parameters is challenging and influences the accuracy of these models. In the present study, the effects of irradiation on the overall flow stress and ductile damage behavior of two ferritic-martensitic steels is parametrically investigated. The results indicate that the ductile damage model parameters are mostly insensitive to irradiation levels at higher dose levels though the resulting flow stress behavior varies significantly.

  13. Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

    Energy Technology Data Exchange (ETDEWEB)

    Byun, Thak Sang, E-mail: byunts@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Toloczko, Mychailo B. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Saleh, Tarik A.; Maloy, Stuart A. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2013-01-15

    To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3-148 dpa at 378-504 Degree-Sign C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 Degree-Sign C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa {radical}m occurred in room temperature tests when irradiation temperature was below 400 Degree-Sign C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa {radical}m was measured when the irradiation temperature was above 430 Degree-Sign C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3-148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 Degree-Sign C) irradiation cases, which indicates that the ductile-brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.

  14. Effect of mechanical restraint on weldability of reduced activation ferritic/martensitic steel thick plates

    Energy Technology Data Exchange (ETDEWEB)

    Serizawa, Hisashi, E-mail: serizawa@jwri.osaka-u.ac.jp [Joining and Welding Research Institute, Osaka University, 11-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Nakamura, Shinichiro [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suite, Osaka 565-0871 (Japan); Tanaka, Manabu; Kawahito, Yousuke [Joining and Welding Research Institute, Osaka University, 11-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Tanigawa, Hiroyasu [Fusion Research and Development Directorate, Japan Atomic Energy Agency, 2-4 Shirakita, Shirane, Naka, Ibaraki 319-1195 (Japan); Katayama, Seiji [Joining and Welding Research Institute, Osaka University, 11-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan)

    2011-10-01

    As one of the reduced activation ferritic/martensitic steels, the weldability of thick F82H plate was experimentally examined using new heat sources in order to minimize the total heat input energy in comparison with TIG welding. A full penetration of 32 mm thick plate could be produced as a combination of a 12 mm deep first layer generated by a 10 kW fiber laser beam and upper layers deposited by a plasma MIG hybrid welding with Ar + 2%O shielding gas. Also, the effect of mechanical restraint on the weldability under EB welding of thick F82H plate was studied by using FEM to select an appropriate specimen size for the basic test. The appropriate and minimum size for the basic test of weldability under EB welding of 90 mm thick plate might be 200 mm in length and 400 mm in width where the welding length should be about 180 mm.

  15. 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.

  16. Radiation induced segregation and precipitation behavior in self-ion irradiated Ferritic/Martensitic HT9 steel

    Science.gov (United States)

    Zheng, Ce; Auger, Maria A.; Moody, Michael P.; Kaoumi, Djamel

    2017-08-01

    In this study, Ferritic/Martensitic (F/M) HT9 steel was irradiated to 20 displacements per atom (dpa) at 600 nm depth at 420 and 440 °C, and to 1, 10 and 20 dpa at 600 nm depth at 470 °C using 5 MeV Fe++ ions. The characterization was conducted using ChemiSTEM and Atom Probe Tomography (APT), with a focus on radiation induced segregation and precipitation. Ni and/or Si segregation at defect sinks (grain boundaries, dislocation lines, carbide/matrix interfaces) together with Ni, Si, Mn rich G-phase precipitation were observed in self-ion irradiated HT9 except in very low dose case (1 dpa at 470 °C). Some G-phase precipitates were found to nucleate heterogeneously at defect sinks where Ni and/or Si segregated. In contrast to what was previously reported in the literature for neutron irradiated HT9, no Cr-rich α‧ phase, χ-phases, η phase and voids were found in self-ion irradiated HT9. The difference of observed microstructures is probably due to the difference of irradiation dose rate between ion irradiation and neutron irradiation. In addition, the average size and number density of G-phase precipitates were found to be sensitive to both irradiation temperature and dose. With the same irradiation dose, the average size of G-phase increased whereas the number density decreased with increasing irradiation temperature. Within the same irradiation temperature, the average size increased with increasing irradiation dose.

  17. Dual and Triple Ion-Beam Irradiations of Fe, Fe(Cr) and Fe(Cr)-ODS Final Report: IAEA SMoRE CRP

    Energy Technology Data Exchange (ETDEWEB)

    Fluss, M J; Hsiung, L L; Marian, J

    2011-11-20

    ., for F82H reduced activation ferritic martensitic (RAF/M) steel. These previous results combined with our data suggest a complex new 'catalytic' mechanism whereby H interacts with the steady state population of defects and the embryonic cavities so as to accelerated cavity (void) growth in both Fe(Cr) and under special conditions in ODS steels.

  18. Fractographic examination of reduced activation ferritic/martensitic steel charpy specimens irradiated to 30 dpa at 370{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S.; Hamilton, M.L. [Pacific Northwest National Lab., Richland, WA (United States); Schubert, L.E. [Univ. of Missouri, Rolla, MO (United States)

    1996-10-01

    Fractographic examinations are reported for a series of reduced activation ferritic/Martensitic steel Charpy impact specimens tested following irradiation to 30 dpa at 370{degrees}C in FFTF. One-third size specimens of six low activation steels developed for potential application as structural materials in fusion reactors were examined. A shift in brittle fracture appearance from cleavage to grain boundary failure was noted with increasing manganese content. The results are interpreted in light of transmutation induced composition changes in a fusion environment.

  19. Radiolysis driven changes to oxide stability during irradiation-corrosion of 316L stainless steel in high temperature water

    Science.gov (United States)

    Raiman, Stephen S.; Bartels, David M.; Was, Gary S.

    2017-09-01

    316L stainless steel samples were irradiated with a proton beam while simultaneously exposed to high temperature water with hydrogen (320 °C, 3 wppm H2, neutral pH) to study the effect of radiation on corrosion. The inner oxides on irradiated samples were found to be depleted in chromium when compared to the inner oxides on unirradiated samples exposed to the same conditions. Additionally, hematite was found on the oxide surfaces of irradiated samples, but not on unirradiated samples. Sample areas which were not directly irradiated but were exposed to the flow of irradiated water also exhibited chromium-deficient inner oxides and had hematite on their surfaces, so it is concluded that water radiolysis is the primary driver of both effects. Thermodynamic calculations and radiolysis modeling were used to show that radiolytic production of hydrogen peroxide was sufficient to raise corrosion potential high enough to cause the dissolution of chromium-rich spinel oxides which make up the inner oxide layer on stainless steel in high temperature water.

  20. HANARO instrumented capsule development for supporting a study on the irradiation damage of stainless steels for nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Y. H.; Cho, M. S.; Sohn, J. M.; Kim, H. R.; Lee, B. C.; Kim, K. H

    2000-10-01

    As a part of the program for the maximum utilization of HANARO by MOST, Korea, an instrumented capsule (00M-01U) was designed and fabricated for supporting a study on the irradiation damage of stainless steels for nuclear applications. The basic structure of the capsule for the irradiation of Stainless steels was based on that of the 99M-01K capsule irradiated successfully in HANARO. To satisfy the user requirements such as irradiation temperature and neutron fluence, the optimal arrangement of test specimens was done in the axial and circumferential direction. The temperature distribution and thermal stress of a capsule with multi-holes were obtained by a finite element analysis code, ANSYS. From these analyzed data, this capsule was found to be compatible with HANARO design requirement. Various types of specimens such as small tensile, Charpy, TEM and EPMA specimens were inserted in the capsule. The specimens will be irradiated in the IR2 test hole of HANARO at 288, 300 and 350 deg C up to a fast neutron fluence of 1.0x10{sup 20}(n/cm{sup 2})(E>1.0MeV)

  1. Irradiation performance of 9--12 Cr ferritic/martensitic stainless steels and their potential for in-core application in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Development of radiation damage during in-situ Kr++ irradiation of Fesbnd Nisbnd Cr model austenitic steels

    Science.gov (United States)

    Desormeaux, M.; Rouxel, B.; Motta, A. T.; Kirk, M.; Bisor, C.; de Carlan, Y.; Legris, A.

    2016-07-01

    In situ irradiations of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti model austenitic steels were performed at the Intermediate Voltage Electron Microscope (IVEM)-Tandem user Facility (Argonne National Laboratory) at 600 °C using 1 MeV Kr++. The experiment was designed in the framework of cladding development for the GEN IV Sodium Fast Reactors (SFR). It is an extension of previous high dose irradiations on those model alloys at JANNuS-Saclay facility in France, aimed at investigating swelling mechanisms and microstructure evolution of these alloys under irradiation [1]. These studies showed a strong influence of Ni in decreasing swelling. In situ irradiations were used to continuously follow the microstructure evolution during irradiation using both diffraction contrast imaging and recording of diffraction patterns. Defect analysis, including defect size, density and nature, was performed to characterize the evolving microstructure and the swelling. Comparison of 15Cr/15Nisbnd Ti and 15Cr/25Nisbnd Ti irradiated microstructure has lent insight into the effect of nickel content in the development of radiation damage caused by heavy ion irradiation. The results are quantified and discussed in this paper.

  3. Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.K.

    2001-01-30

    A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

  4. Raman spectroscopic analysis of iron chromium oxide microspheres generated by nanosecond pulsed laser irradiation on stainless steel.

    Science.gov (United States)

    Ortiz-Morales, M; Soto-Bernal, J J; Frausto-Reyes, C; Acosta-Ortiz, S E; Gonzalez-Mota, R; Rosales-Candelas, I

    2015-06-15

    Iron chromium oxide microspheres were generated by pulsed laser irradiation on the surface of two commercial samples of stainless steel at room temperature. An Ytterbium pulsed fiber laser was used for this purpose. Raman spectroscopy was used for the characterization of the microspheres, whose size was found to be about 0.2-1.7 μm, as revealed by SEM analysis. The laser irradiation on the surface of the stainless steel modified the composition of the microspheres generated, affecting the concentration of the main elemental components when laser power was increased. Furthermore, the peak ratio of the main bands in the Raman spectra has been associated to the concentration percentage of the main components of the samples, as revealed by Energy-Dispersive X-ray Spectroscopy (EDS) analysis. These experiments showed that it is possible to generate iron chromium oxide microspheres on stainless steel by laser irradiation and that the concentration percentage of their main components is associated with the laser power applied.

  5. Swelling and microstructure of austenitic stainless steel ChS-68 CW after high dose neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Konobeev, Yu.V. [State Scientific Center of Russian Federation - Institute of Physics and Power Engineering (IPPE), Obninsk, Kaluga Region (Russian Federation); Garner, F.A., E-mail: frank.garner@dslextreme.co [Radiation Effects Consulting, 2003 Howell Avenue, Richland, WA 99354 (United States)

    2009-08-15

    Austenitic stainless steel ChS-68 serving as fuel pin cladding was irradiated in the 20% cold-worked condition in the BN-600 fast reactor in the range 56-84 dpa. This steel was developed to replace EI-847 which was limited by its insufficient resistance to void swelling. Comparison of swelling between EI-847 and ChS-68 under similar irradiation conditions showed improvement of the latter steel by an extended transient regime of an additional approx10 dpa. Concurrent with swelling was the development of a variety of phases. In the temperature range 430-460 deg. S where the temperature peak of swelling was located, the principal type of phase generated during irradiation was G-phase, with volume fraction increasing linearly with dose to approx0.5% at 84 dpa. While the onset of swelling is concurrent with formation of G-phase, the action of G-phase cannot be confidently ascribed to significant removal from solution of swelling-suppressive elements such as silicon. A plausible mechanism for the higher resistance to void swelling of ChS-68 as compared with EI-847 may be related to an observed higher stability of faulted dislocation loops in ChS-68 that impedes the formation of a glissile dislocation network. The higher level of boron in ChS-68 is thought to be one contributor that might play this role.

  6. The effect of low dose rate irradiation on the swelling of 12% cold-worked 316 stainless steel.

    Energy Technology Data Exchange (ETDEWEB)

    Allen, T. R.

    1999-03-02

    In pressurized water reactors (PWRs), stainless steel components are irradiated at temperatures that may reach 400 C due to gamma heating. If large amounts of swelling (>10%) occur in these reactor internals, significant swelling related embrittlement may occur. Although fast reactor studies indicate that swelling should be insignificant at PWR temperatures, the low dose rate conditions experienced by PWR components may possibly lead to significant swelling. To address these issues, JNC and ANL have collaborated to analyze swelling in 316 stainless steel, irradiated in the EBR-II reactor at temperatures from 376-444 C, at dose rates between 4.9 x 10{sup {minus}8} and 5.8 x 10{sup {minus}7} dpa/s, and to doses of 56 dpa. For these irradiation conditions, the swelling decreases markedly at temperatures less than approximately 386 C, with the extrapolated swelling at 100 dpa being around 3%. For temperatures greater than 386 C, the swelling extrapolated to 100 dpa is around 9%. For a factor of two difference in dose rate, no statistically significant effect of dose rate on swelling was seen. For the range of dose rates analyzed, the swelling measurements do not support significant (>10%) swelling of 316 stainless steel in PWRs.

  7. Mechanical properties of neutron-irradiated nickel-containing martensitic steels: II. Review and analysis of helium-effects studies

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States)]. E-mail: kluehrl@ornl.gov; Hashimoto, N. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Sokolov, M.A. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Maziasz, P.J. [Oak Ridge National Laboratory, Metals and Ceramics Division, Building 4500S, P.O. Box 2008, MS 6151, Oak Ridge, TN 37831-6151 (United States); Shiba, K. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan); Jitsukawa, S. [Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Tokai, Ibaraki 319-1195 (Japan)

    2006-10-15

    In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr-1MoVNb (modified 9Cr-1Mo) and 12Cr-1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10-12 dpa at 300 and 400 deg. C and in the Fast Flux Test Facility (FFTF) to 15 dpa at 393 deg. C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed.

  8. Integrated analysis of millisecond laser irradiation of steel by comprehensive optical diagnostics and numerical simulation

    Science.gov (United States)

    Doubenskaia, M.; Smurov, I.; Nagulin, K. Yu.

    2016-04-01

    Complimentary optical diagnostic tools are applied to provide comprehensive analysis of thermal phenomena in millisecond Nd:YAG laser irradiation of steel substrates. The following optical devices are employed: (a) infrared camera FLIR Phoenix RDASTM equipped by InSb sensor with 3 to 5 µm band pass arranged on 320 × 256 pixels array, (b) ultra-rapid camera Phantom V7.1 with SR-CMOS monochrome sensor in the visible spectral range, up to 105 frames per second for 64 × 88 pixels array, (c) original multi-wavelength pyrometer in the near-infrared range (1.370-1.531 µm). The following laser radiation parameters are applied: variation of energy per pulse in the range 15-30 J at a constant pulse duration of 10 ms with and without application of protective gas (Ar). The evolution of true temperature is restored based on the method of multi-colour pyrometry; by this way, melting/solidification dynamics is analysed. Emissivity variation with temperature is studied, and hysteresis type functional dependence is found. Variation of intensity of surface evaporation visualised by the camera Phantom V7.1 is registered and linked with the surface temperature evolution, different surface roughness and influence of protective gas atmosphere. Determination of the vapour plume temperature based on relatively intensities of spectral lines is done. The numerical simulation is carried out applying the thermal model with phase transitions taken into account.

  9. Phase diffusionless γ↔α transformations and their effect on physical, mechanical and corrosion properties of austenitic stainless steels irradiated with neutrons and charged particles

    Science.gov (United States)

    Maksimkin, O. P.

    2016-04-01

    The work presents relationships of γ→α' and α'→γ-transformations in reactor 12Cr18Ni10Ti and 08Cr16Ni11Mo3 austenitic stainless steels induced by cold work, irradiation and/or temperature. Energy and mechanical parameters of nucleation and development of deformation-induced martensitic α'-phase in the non-irradiated and irradiated steels are given. The mechanisms of localized static deformation were investigated and its effect on martensitic γ→α' transformation is determined. It has been shown that irradiation of 12Cr18Ni10Ti steel with heavy Kr ions (1.56MeV/nucleon, fluence of 1·1015 cm-2) results in formation of α'-martensite in near-surface layer of the sample. Results of systematic research on reversed α'→γ-transformation in austenitic metastable stainless steels irradiated with slow (VVR-K) and fast (BN-350) neutrons are presented. The effect of annealing on strength and magnetic characteristics was determined. It was found that at the temperature of 400 °C in the irradiated with neutrons samples (59 dpa) an increase of ferromagnetic α'-phase and microhardness was observed. The obtained results could be used during assessment of operational characteristics of highly irradiated austenitic steels during transportation and storage of Fuel Assemblies for fast nuclear reactors.

  10. Effects of heavy-ion irradiation on solute segregation to dislocations in oxide-dispersion-strengthened Eurofer 97 steel

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Ceri A., E-mail: ceri.williams@materials.ox.ac.uk [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); Hyde, Jonathan M. [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom); National Nuclear Laboratory, B168 Harwell, Didcot, Oxon OX11 0QT (United Kingdom); Smith, George D.W.; Marquis, Emmanuelle A. [Department of Materials, University of Oxford, Parks Road, Oxford, OX1 3PH (United Kingdom)

    2011-05-01

    Interactions of solute atoms with dislocations are known to influence the mechanical properties of metals, and are an important aspect of irradiated materials. Here, atom-probe tomography is used to investigate the segregation of solutes to dislocations in the martensitic/ferritic oxide-dispersion-strengthened Eurofer steel. The material was irradiated with 0.5-2 MeV Fe{sup 2+} ions at 400 deg. C, and the distributions of solutes at dislocations were analysed in the 'as-received' condition, after annealing, and after ion implantation. While Mn is randomly distributed in the as-received material, significant segregation of Mn at dislocations and grain boundaries is observed after irradiation. The level of Mn segregation varies depending on the density of oxide particles surrounding the dislocations.

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

    Science.gov (United States)

    Subbotin, A. V.; Panyukov, S. V.

    2016-08-01

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

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

    CERN Document Server

    Subbotina, A V

    2016-01-01

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. We show that the obtained results are in a good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. 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.

  13. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Science.gov (United States)

    Chiapetto, M.; Becquart, C. S.; Domain, C.; Malerba, L.

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a "grey alloy" approach that extends the already existing OKMC model for neutron irradiated Fe-C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe-C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  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. Structural and mechanical properties of welded joints of reduced activation martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Filacchioni, G. E-mail: gianni.filacchioni@casaccia.enea.it; Montanari, R.; Tata, M.E.; Pilloni, L

    2002-12-01

    Gas tungsten arc welding and electron beam welding methods were used to realise welding pools on plates of reduced activation martensitic steels. Structural and mechanical features of these simulated joints have been investigated in as-welded and post-welding heat-treated conditions. The research allowed to assess how each welding technique affects the original mechanical properties of materials and to find suitable post-welding heat treatments. This paper reports results from experimental activities on BATMAN II and F82H mod. steels carried out in the frame of the European Blanket Project - Structural Materials Program.

  16. Accumulation and annealing of radiation defects under low-temperature electron and neutron irradiation of ODS steel and Fe-Cr alloys

    Science.gov (United States)

    Arbuzov, V. L.; Goshchitskii, B. N.; Sagaradze, V. V.; Danilov, S. E.; Kar'kin, A. E.

    2010-10-01

    The processes of accumulation and annealing of radiation defects at low-temperature (77 K) electron and neutron irradiation and their effect on the physicomechanical properties of Fe-Cr alloys and oxide dispersion strengthened (ODS) steel have been studied. It has been shown that the behavior of radiation defects in ODS steel and Fe-Cr alloys is qualitatively similar. Above 250 K, radiation-induced processes of the solid solution decomposition become conspicuous. These processes are much less pronounced in ODS steel because of specific features of its microstructure. Processes related to the overlapping of displacement cascades under neutron irradiation have been considered. It has been shown that, in this case, it is the increase in the size of vacancy clusters, rather than the growth of their concentration, that is prevailing. Possible mechanisms of the radiation hardening of the ODS steel and the Fe-13Cr alloy upon irradiation and subsequent annealing have been discussed.

  17. Compatibility tests on steels in molten lead and lead-bismuth

    Energy Technology Data Exchange (ETDEWEB)

    Fazio, C. E-mail: concetta@netbra.brasimone.enea.it; Benamati, G.; Martini, C.; Palombarini, G

    2001-07-01

    The compatibility of steels with liquid lead and liquid lead-bismuth is a critical issue for the development of accelerator-driven system (ADS). In this work the results of a set of preliminary tests carried out in stagnant molten lead at 737 K and in lead-bismuth at 573, 673 and 749 K are summarised. The tests were conducted for 700, 1200, 1500 and 5000 h. Three steels were tested: two martensitic steels (mod. F82H and MANET II) and one austenitic steel (AISI 316L). The martensitic steels underwent oxidation phenomena at the higher testing temperature, due to oxygen dissolved in the melts. At a lower test temperature (573 K) and higher exposure time (5000 h) the oxidation rate of the martensitic steel seems to be lower and the developed oxide layer protective against liquid metal corrosion. The austenitic steel, in turn, exhibited an acceptable resistance to corrosion-oxidation under the test conditions.

  18. Helium effects on microstructural change in RAFM steel under irradiation: Reaction rate theory modeling

    Science.gov (United States)

    Watanabe, Y.; Morishita, K.; Nakasuji, T.; Ando, M.; Tanigawa, H.

    2015-06-01

    Reaction rate theory analysis has been conducted to investigate helium effects on the formation kinetics of interstitial type dislocation loops (I-loops) and helium bubbles in reduced-activation-ferritic/martensitic steel during irradiation, by focusing on the nucleation and growth processes of the defect clusters. The rate theory model employs the size and chemical composition dependence of thermal dissociation of point defects from defect clusters. In the calculations, the temperature and the production rate of Frenkel pairs are fixed to be T = 723 K and PV = 10-6 dpa/s, respectively. And then, only the production rate of helium atoms was changed into the following three cases: PHe = 0, 10-7 and 10-5 appm He/s. The calculation results show that helium effect on I-loop formation quite differs from that on bubble formation. As to I-loops, the loop formation hardly depends on the existence of helium, where the number density of I-loops is almost the same for the three cases of PHe. This is because helium atoms trapped in vacancies are easily emitted into the matrix due to the recombination between the vacancies and SIAs, which induces no pronounced increase or decrease of vacancies and SIAs in the matrix, leading to no remarkable impact on the I-loop nucleation. On the other hand, the bubble formation depends much on the existence of helium, in which the number density of bubbles for PHe = 10-7 and 10-5 appm He/s is much higher than that for PHe = 0. This is because helium atoms trapped in a bubble increase the vacancy binding energy, and suppress the vacancy dissociation from the bubble, resulting in a promotion of the bubble nucleation. And then, the helium effect on the promotion of bubble nucleation is very strong, even the number of helium atoms in a bubble is not so large.

  19. Neutron-irradiation + helium hardening and embrittlement modeling of 9% Cr-steels in an engineering perspective (HELENA)

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, Rachid

    2008-07-01

    This report provides a physically-based engineering model to estimate the radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades. However, a number of assumptions and simplifications were considered to obtain a simple description of irradiation hardening and embrittlement primarily relying on the available experimental data. As a result, two components were basically identified, the dpa component that can be associated with black dots and small loops and the He-component accounting for helium bubbles. The dpa component is strongly dependent on the irradiation temperature and its dependence law was based on a first-order annealing kinetics. The damage accumulation law was also modified to take saturation into account. Finally, the global kinetics of the damage accumulation kept defined, its amplitude is fitted to one experimental condition. The model was rationalized on an experimental database that mainly consists of {proportional_to}9%Cr-steels irradiated in the technologically important temperature range of 50 to 600 C up do 50 dpa and with a He-content up to {proportional_to}5000 appm, including neutron and proton irradiation as well as implantation. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Finally, the hardening-to-embrittlement correlation is obtained using the load diagram approach. Despite the large experimental scatter, inherent to the variety of the materials and irradiation as well as testing conditions, the obtained results are very promising. Improvement of the model performance is still possible by including He-hardening saturation and high temperature softening but unfortunately, at this stage, a number of conflicting experimental data

  20. Effects of helium pre-implantation on the microstructure and mechanical properties of irradiated 316 stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B.; Tedeski, G.R.; Lucas, G.E.; Odette, G.R. [Univ. of California, Santa Barbara, CA (United States). Dept. of Chemical and Nuclear Engineering; Stoller, R.E. [Oak Ridge National Lab., TN (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-11-01

    Transmission electron microscopy (TEM) specimens of a First Core heat of 316 stainless steel, in both the solution annealed and 20% cold worked condition, were irradiated to 46 dpa at 420 C, to 49 dpa at 520 C, and to 34 dpa at 600 C in FFTF/MOTA. Prior to irradiation, about half of the specimens were pre-implanted with approximately 100 appm of helium, and of these, several of the solution annealed and pre-implanted specimens were aged at 800 C for 2 hr. Post-irradiation density measurements showed little difference in density between the unimplanted alloys and their helium implanted counterparts. Microstructural observations on specimens irradiated at 420 C and 520 C showed relatively minor differences in defect distributions between the unimplanted and the helium implanted materials; in all cases the defect sizes and number densities were consistent with data in the literature. Where possible, irradiation hardening of the alloys was experimentally evaluated by microhardness and shear punch; experimentally obtained values were compared to values calculated using a computer model based on barrier hardening and the microstructural data. All methods indicated relatively small effects of helium implantation, and both measured and calculated values were in agreement with the range of values reported in the literature.

  1. Ion-irradiation effects on dissimilar friction stir welded joints between ODS alloy and ferritic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.-L., E-mail: chunliang@mail.ndhu.edu.tw [Department of Materials Science and Engineering, National Dong-Hwa University, Hualien 97401, Taiwan (China); Richter, A. [Department of Engineering, Technical University of Applied Sciences Wildau, Bahnhofstrasse 1, 15745 Wildau (Germany); Kögler, R. [Institute of Ion Beam Physics and Materials Research, Helmholtz Center Dresden-Rossendorf (HZDR), Bautzner Landstraße 400, 01328 Dresden (Germany); Griepentrog, M.; Reinstädt, P. [BAM Federal Institute for Materials Research and Testing, Unter den Eichen 87, 12205 Berlin (Germany)

    2014-12-05

    Highlights: • FSSW has successfully been used in the welding of dissimilar materials. • The irradiation causes different degrees of hardening in the welding zones. • The formation of He bubbles at precipitates was found in the dissimilar joints. • The hardening effect is due to formation of He-filled vacancies. - Abstract: Friction stir spot welding (FSSW) is an advanced technique for the joining of materials to prevent agglomeration of fine oxide particles, grain coarsening, and stress corrosion cracking etc. In this study, the dissimilar FSSW joint of stainless steel 430/ODS was irradiated with a Fe{sup +}/He{sup +} dual ion beam. Irradiation damage can cause deterioration in the mechanical properties especially in the welding zones. The joint quality therefore plays a decisive role in the life expectancy of nuclear reactors. The effect of irradiation on different zones in the joint (the thermo-mechanically affected zone, the heat affected zone and the base material) was investigated by TEM and nanoindentation. Irradiation causes a hardness increase in all welding zones with a characteristic hardness maximum. The relative hardness increase and the related microstructure are discussed. The formation of He bubbles at chromium carbide precipitates and the homogeneous distribution of He filled vacancies in the mixture region of the 430/ODS FSSW joints was observed.

  2. 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)

  3. Investigation on femto-second laser irradiation assisted shock peening of medium carbon (0.4% C) steel

    Science.gov (United States)

    Majumdar, Jyotsna Dutta; Gurevich, Evgeny L.; Kumari, Renu; Ostendorf, Andreas

    2016-02-01

    In the present study, the effect of femtosecond laser irradiation on the peening behavior of 0.4% C steel has been evaluated. Laser irradiation has been conducted with a 100 μJ and 300 fs laser with multiple pulses under varied energy. Followed by laser irradiation, a detailed characterization of the processed zone was undertaken by scanning electron microscopy, and X-ray diffraction technique. Finally, the residual stress distribution, microhardness and wear resistance properties of the processed zone were also evaluated. Laser processing leads to shock peening associated with plasma formation and its expansion, formation of martensite and ferrito-pearlitic phase in the microstructure. Due to laser processing, there is introduction of residual stress on the surface which varies from high tensile (140 MPa) to compressive (-335 MPa) as compared to 152 MPa of the substrate. There is a significant increase in microhardness to 350-500 VHN as compared to 250 VHN of substrate. The fretting wear behavior against hardened steel ball shows a significant reduction in wear depth due to laser processing. Finally, a conclusion of the mechanism of wear has been established.

  4. Hardness and microstructural response to thermal annealing of irradiated ASTM A533B class 1 plate steel

    Energy Technology Data Exchange (ETDEWEB)

    Reinhart, D.E. [SMS Concast, Inc., Pittsburgh, PA (United States); Kumar, A.S. [Univ. of Missouri, Rolla, MO (United States); Gelles, D.S.; Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States); Rosinski, S.T. [Electric Power Research Inst., Charlotte, NC (United States)

    1999-10-01

    Hardness measurements were used to determine the post-irradiation annealing response of A533B class 1 plate steel irradiated to a fluence of 1 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV) at 150 C. Rockwell hardness measurements indicated that the material had hardened by 6.6 points on the B scale after irradiation. The irradiation induced hardness increase was associated with a decrease in upper shelf energy from 63.4 J to 5-1.8 J and a temperature shift in the Charpy curve at the 41 J level from 115 C to 215 C. Specimens were annealed after irradiation at temperatures of 343 C (650 F), 399 C (750 F), and 454 C (850 F) for durations of up to one week (168 h). Hardness measurements were made to chart recovery of hardness as a function of time and temperature. Specimens annealed at the highest temperature 454 C recovered the fastest, fully recovering within 144 h. Specimens annealed at 399 C recovered completely within 168 h. Specimens annealed at the lowest temperature, 343 C recovered only {approximately}70% after 168 h of annealing. After neutron irradiation, a new feature of black spot damage was found to be superimposed on the unirradiated microstructure. The density of black spots was found to vary from 2.3 {times} 10{sup 15}/cm{sup 3} to 1.1 {times} 10{sup 16}/cm{sup 3} with an average diameter of 2.85 nm. Following annealing at 454 C for 24 h the black spot damage was completely annealed out. It was concluded that the black spot damage was responsible for 70% of the irradiation-induced hardness.

  5. Modelling of the plasticity and brittle failure of the irradiated bainitic steels; Modelisation du comportement en plasticite et a rupture des aciers bainitiques irradies

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, C.N.

    2010-02-15

    Low alloy steels are used in various equipments of nuclear reactors. Subjected to neutron irradiation produced during the operation of reactors, these materials exhibit significant changes in their microstructure, especially with the formation of radiation defects as interstitial loops, void clusters and precipitates. These defects in interactions with dislocations lead to a hardening and embrittlement which are directly related to the received dose and neutron flux. The plastic behaviour of non-irradiated low alloy bainitic steels has been the object of several modelling based on observations from experiments and atomistic simulations. Some of them result from thesis supported by EDF and CEA, which describe different strategies for the micro-mechanical modelling of brittle failure. Improvements in this work come from the integration of new physical characteristics and the attention paid to the representativeness of the microstructure: whereas realistic microstructures in terms of morphology and crystal orientations have been adopted, a dislocation density based constitutive model in the large deformation framework is used to describe crystal plasticity. This choice is justified by the need to take into account, in the constitutive modelling, the interactions between dislocations and irradiation defects under severe loading conditions. The plasticity laws have been implemented in the finite elements code ZeBuLoN in order to perform computations of polycrystalline aggregates. Such aggregates are representative volume elements. They thus provide the database required for the application of brittle failure models to structures. This multi-scale character confers to the modelling the status of 'micro-mechanical local approach of failure'. (author)

  6. Small Punch Test on Before and Post Irradiated Domestic Reactor Pressure Steel

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    Problems may be caused when applying the standard specimen to study the properties of irradiated reactor materials, because of its big dimension, e.g.: The inner temperature gradient of the specimen is high when irradiated, the radiation

  7. The formation of radiation-induced segregation at twin bands in ion-irradiated austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hyung-Ha; Lee, Gyeong-Geun; Kwon, Junhyun; Hwang, Seong Sik [Nuclear Materials Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Shin, Chansun, E-mail: c.shin@mju.ac.kr [Department of Materials Science and Engineering, Myongji University, 116 Myongji-ro, Cheoin-gu, Youngin, Gyeonggi-do 449-728 (Korea, Republic of)

    2014-11-15

    Radiation-induced segregation (RIS) at twins was investigated using transmission electron microscopy (TEM) for ion-irradiated austenitic stainless steel. Significant RIS was found to occur at twin boundaries. TEM analysis indicates that interfacial dislocations at partially coherent twin boundaries are potential sites for strong RIS phenomenon. The RIS causes the formation of thin bands having a higher Ni and lower Cr concentration in twin bands with a width less than 15 nm. In wider twin bands, strong RIS occurs only at the outer twin boundaries, but not inside the band. The possible mechanism for the formation of the RIS thin band is discussed.

  8. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. C.; Lee, B. S. [KAERI, Taejon (Korea, Republic of); Oh, Y. J. [Hanbat National Univ., Taejon (Korea, Republic of)

    2003-10-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T{sub sp}), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T{sub 41J}. T{sub sp} from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T{sub 0}) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  9. Stability of the strengthening nanoprecipitates in reduced activation ferritic steels under Fe{sup 2+} ion irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Tan, L., E-mail: tanl@ornl.gov; Katoh, Y.; Snead, L.L.

    2014-02-01

    The stability of MX-type precipitates is critical to retain mechanical properties of both reduced activation ferritic–martensitic (RAFM) and conventional FM steels at elevated temperatures. Radiation resistance of TaC, TaN, and VN nanoprecipitates irradiated up to ∼49 dpa at 500 °C using Fe{sup 2+} is investigated in this work. Transmission electron microscopy (TEM) utilized in standard and scanning mode (STEM) reveals the non-stoichiometric nature of the nanoprecipitates. Irradiation did not alter their crystalline nature. The radiation resistance of these precipitates, in an order of reduced resistance, is TaC, VN, and TaN. Particle dissolution, growth, and reprecipitation were the modes of irradiation-induced instability. Irradiation also facilitated formation of Fe{sub 2}W type Laves phase limited to the VN and TaN bearing alloys. This result suggests that nitrogen level should be controlled to a minimal level in alloys to gain greater radiation resistance of the MX-type precipitates at similar temperatures as well as postpone the formation and subsequent coarsening of Laves phase.

  10. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels; Programas de investigacion sobre fragilizacion por irradiacion de los aceros de vasija

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F. [Ciemat. Madrid (Spain)

    2000-07-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study the evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

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

    Science.gov (United States)

    Kasahara, Shigeki; Kitsunai, Yuji; Chimi, Yasuhiro; Chatani, Kazuhiro; Koshiishi, Masato; Nishiyama, Yutaka

    2016-11-01

    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.

  12. Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

    Science.gov (United States)

    Böhmert, J.; Viehrig, H.-W.; Ulbricht, A.

    2004-08-01

    Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.

  13. The changes of the structural, magnetic, and mechanical properties in a reactor pressure vessel steel neutron-irradiated at 70 .deg. C

    CERN Document Server

    Park, D G; Jang, K S; Jung, M M; Kim, G M

    1999-01-01

    The irradiation embrittlement of reactor-pressure-vessel steel has been one of the main safety concerns in nuclear power plants. In the present study, an SA508-3 RPV steel was irradiated by neutrons with various fluences up to 10 sup 1 sup 8 n/cm sup 2 (E>=1MeV) at a temperature of approximately 70 .deg. C. The irradiation responses of the structural, the magnetic, and the mechanical properties of the steel were investigated by means of X-ray diffraction, Moessbauer spectroscopy, magnetic Barkhausen noise, and micro-Vickers hardness measurements. The transitions of all of these parameters occurred above a neutron does of 10 sup 1 sup 6 n/cm sup 2. The results of the X-ray and the Moessbauer experiments revealed that neutron irradiation led to the possibility of partial amorphization in the investigated RPV steel. The changes of the physical and the mechanical properties were discussed in terms of irradiation-induced cascade damage of crystalline materials.

  14. New Dosimetric Interpretation of the DV50 Vessel-Steel Experiment Irradiated in the OSIRIS MTR Reactor Using the Monte-Carlo Code TRIPOLI-4®

    Directory of Open Access Journals (Sweden)

    Malouch Fadhel

    2016-01-01

    Full Text Available An irradiation program DV50 was carried out from 2002 to 2006 in the OSIRIS material testing reactor (CEA-Saclay center to assess the pressure vessel steel toughness curve for a fast neutron fluence (E > 1 MeV equivalent to a French 900-MWe PWR lifetime of 50 years. This program allowed the irradiation of 120 specimens out of vessel steel, subdivided in two successive irradiations DV50 n∘1 and DV50 n∘2. To measure the fast neutron fluence (E > 1 MeV received by specimens after each irradiation, sample holders were equipped with activation foils that were withdrawn at the end of irradiation for activity counting and processing. The fast effective cross-sections used in the dosimeter processing were determined with a specific calculation scheme based on the Monte-Carlo code TRIPOLI-3 (and the nuclear data ENDF/B-VI and IRDF-90. In order to put vessel-steel experiments at the same standard, a new dosimetric interpretation of the DV50 experiment has been performed by using the Monte-Carlo code TRIPOLI-4 and more recent nuclear data (JEFF3.1.1 and IRDF-2002. This paper presents a comparison of previous and recent calculations performed for the DV50 vessel-steel experiment to assess the impact on the dosimetric interpretation.

  15. Extreme embrittlement of austenitic stainless steel irradiated to 75--81 dpa at 335--360 C

    Energy Technology Data Exchange (ETDEWEB)

    Porollo, S.I.; Vorobjev, A.N.; Konobeev, Yu.V. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States)

    1998-03-01

    This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0--2000 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approximately}270 C. Tubes in the annealed condition reached 75 dpa at 335 C, and another set in the 20% cold-worked condition reached 81 dpa at 360 C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes. The embrittlement is explained in terms of the sensitivity of the swelling regime to displacement rate and the large, unprecedented levels of swelling reached at 335--360 C at these high neutron fluences. The failure mechanism appears to be identical to that observed at similar swelling levels in other austenitic steels irradiated in US fast reactors at 400--425 C, whereby stress-concentration between voids and nickel segregation at void surfaces predisposes the steel to an epsilon martensite transformation followed by formation of alpha martensite at crack tips. The very slow strain rate inherent in such creep tests and the relatively high helium levels may also contribute to the failure.

  16. Cluster dynamics modeling of the effect of high dose irradiation and helium on the microstructure of austenitic stainless steels

    Science.gov (United States)

    Brimbal, Daniel; Fournier, Lionel; Barbu, Alain

    2016-01-01

    A mean field cluster dynamics model has been developed in order to study the effect of high dose irradiation and helium on the microstructural evolution of metals. In this model, self-interstitial clusters, stacking-fault tetrahedra and helium-vacancy clusters are taken into account, in a configuration well adapted to austenitic stainless steels. For small helium-vacancy cluster sizes, the densities of each small cluster are calculated. However, for large sizes, only the mean number of helium atoms per cluster size is calculated. This aspect allows us to calculate the evolution of the microstructural features up to high irradiation doses in a few minutes. It is shown that the presence of stacking-fault tetrahedra notably reduces cavity sizes below 400 °C, but they have little influence on the microstructure above this temperature. The binding energies of vacancies to cavities are calculated using a new method essentially based on ab initio data. It is shown that helium has little effect on the cavity microstructure at 300 °C. However, at higher temperatures, even small helium production rates such as those typical of sodium-fast-reactors induce a notable increase in cavity density compared to an irradiation without helium.

  17. In situ TEM study of G-phase precipitates under heavy ion irradiation in CF8 cast austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Wei-Ying [Argonne National Laboratory, Argonne, IL 60439 (United States); University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Li, Meimei; Zhang, Xuan; Kirk, Marquis A.; Baldo, Peter M. [Argonne National Laboratory, Argonne, IL 60439 (United States); Lian, Tiangan [Electric Power Research Institute, Palo Alto, CA 94304 (United States)

    2015-09-15

    Thermally-aged cast austenitic stainless steels (CASS) CF8 was irradiated with 1 MeV Kr ions at 300, 350 and 400 °C to 1.88 × 10{sup 19} ions/m{sup 2} (∼3 dpa) at the IVEM-Tandem Facility at the Argonne National Laboratory. Before irradiation, the distribution of G-phase precipitates in the ferrite showed spatial variations, and both their size and density were affected by the ferrite–austenite phase boundary and presence of M{sub 23}C{sub 6} carbides. Under 300 °C irradiation, in situ TEM observation showed G-phase precipitates were relatively unchanged in the vicinity of the phase boundary M{sub 23}C{sub 6} carbides, while the density of G-phase precipitates increased with increasing dose within the ferrite matrix. Coarsening of G-phase precipitates was observed in the vicinity of phase boundary M{sub 23}C{sub 6} carbides at 350 °C and 400 °C.

  18. Evolution of microstructure and mechanical properties of VVER-1000 RPV steels under re-irradiation

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Erak, D.; Zhurko, D.

    2015-01-01

    This is a comprehensive study of microstructure and mechanical properties evolution at re-irradiation after recovery annealing of VVER-1000 RPV weld and base metals as well as the effect of annealing on the microstructure and properties of base metal in the zone of the temperature gradient that is implemented during annealing using special heating device. It is shown that the level of radiation-induced microstructural changes under accelerated re-irradiation of weld and base metal is not higher than for the primary irradiation. Thus, we can predict that re-embrittlement of VVER-1000 RPV materials considering the flux effect will not exceed the typical embrittlement rate for the primary irradiation.

  19. Irradiation Test of RPV Steel in 49-2 Research Reactor

    Institute of Scientific and Technical Information of China (English)

    LIN; Yun; TONG; Zhen-feng; ZHANG; Chang-yi; NING; Guang-sheng; YANG; Wen

    2013-01-01

    Reactor pressure vessel(RPV)is the critical and unchangeable component of the PWR during its service lifetime which determines the lifetime of nuclear power plant(NPP).The property decline of RPVis mainly caused by large dose and high energy neutron irradiation during the long term service.Irradiation can reduce the toughness and increase the brittleness of the RPV materials.Thus raising the

  20. Evaluation of ductile-brittle transition temperature before and after neutron irradiation for RPV steels using small punch tests

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Chul [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)]. E-mail: mckim@kaeri.re.kr; Oh, Yong Jun [Hanbat National University, Deogmyeong-dong, Yuseong-gu, Daejeon 305-719 (Korea, Republic of); Lee, Bong Sang [Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2005-08-01

    Small punch (SP) tests were performed to evaluate the ductile-brittle transition temperature before and after a neutron irradiation of reactor pressure vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the conventional Charpy tests and the Master Curve fracture toughness tests in accordance with the American Society for Testing and Materials (ASTM) standard E1921. Small punch specimens were taken from a 1/4t location of the vessel thickness and machined into a 10 mm x 10 mm x 0.5 mm dimension. The specimens were irradiated in the research reactors at Korea Atomic Energy Research Institute Nuclear Research Institute in the Czech Republic at the different fluence levels of about 290 deg C. Small punch tests were performed in the temperature range of RT to -196 deg C using a 2.4 mm diameter ball. For the materials before and after irradiation, the small punch transition temperatures (T {sub SP}), which are determined at the middle of the upper small punch energies, showed a linear correlation with the Charpy index temperature, T {sub 41J}. T {sub SP} from the irradiated samples was increased with the fluence levels and was well within the deviation range of the unirradiated data. However, the transition temperature shift from the Charpy test ({delta}T {sub 41J}) shows a better correlation with the transition temperature shift ({delta}T {sub SP(E)}) when a specific small punch energy level rather than the middle energy level of the small punch curve is used to determine the transition temperature. T {sub SP} also had a correlation with the reference temperature (T {sub 0}) from the Master Curve method using a pre-cracked Charpy V-notched (PCVN) specimen.

  1. Microstructural evolution of austenitic stainless steels irradiated to 17 dpa in spectrally tailored experiment of the ORR and HFIR at 400{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, E.; Hashimoto, N.; Gibson, L.T. [Oak Ridge National Lab., TN (United States)] [and others

    1997-08-01

    The microstructural evolution of austenitic JPCA aged and solution annealed JPCA, 316R, C, K, and HP steels irradiated at 400{degrees}C in spectrally tailored experiments of the ORR and HFIR has been investigated. The helium generation rates were about 12-16 appm He/dpa on the average up to 17.3 dpa. The number densities and average diameters of dislocation loops in the steels have ranges of 3.3 x 10{sup 21} m{sup -3} and 15.2-26.3 nm, respectively, except for HP steel for which they are 1.1 x 10{sup 23} m{sup -3} and 8.0 nm. Precipitates are formed in all steels except for HP steel, and the number densities and average diameters have ranges of 5.2 x 10{sup 20} - 7.7 x 10{sup 21} m{sup -3} and 3.4- 19.3 nm, respectively. In the 216R, C, and K steels, the precipitates are also formed at grain boundaries, and the mean sizes of these are about 110, 50, and 50 nm, respectively. The number densities of cavities are about 1 x 10{sup 22} m{sup -3} in all the steels. The swelling is low in the steels which form the precipitates.

  2. Neutron irradiation and high temperature effects on amorphous Fe-based nano-coatings on steel - A macroscopic assessment

    Science.gov (United States)

    Simos, N.; Zhong, Z.; Dooryhee, E.; Ghose, S.; Gill, S.; Camino, F.; Şavklıyıldız, İ.; Akdoğan, E. K.

    2017-06-01

    The study revealed that loss of ductility in an amorphous Fe-alloy coating on a steel substrate composite structure was essentially prevented from occurring, following radiation with modest neutron doses of ∼2 × 1018 n/cm2. At the higher neutron dose of ∼2 × 1019, macroscopic stress-strain analysis showed that the amorphous Fe-alloy nanostructured coating, while still amorphous, experienced radiation-induced embrittlement, no longer offering protection against ductility loss in the coating-substrate composite structure. Neutron irradiation in a corrosive environment revealed exemplary oxidation/corrosion resistance of the amorphous Fe-alloy coating, which is attributed to the formation of the Fe2B phase in the coating. To establish the impact of elevated temperatures on the amorphous-to-crystalline transition in the amorphous Fe-alloy, electron microscopy was carried out which confirmed the radiation-induced suppression of crystallization in the amorphous Fe-alloy nanostructured coating.

  3. Impact of nuclear irradiation on helium bubble nucleation at interfaces in liquid metals coupled to permeation through stainless steels

    CERN Document Server

    Fradera, Jorge

    2013-01-01

    The impact of nucleating gas bubbles in the form of a dispersed gas phase on hydrogen isotope permeation at interfaces between liquid metals, like LLE, and structural materials, like stainless steel, has been studied. Liquid metal to structural material interfaces involving surfaces, may lower the nucleation barrier promoting bubble nucleation at active sites. Hence, hydrogen isotope absorption into gas bubbles modelling and control at interfaces may have a capital importance regarding design, operation and safety. He bubbles as a permeation barrier principle is analysed showing a significant impact on hydrogen isotope permeation, which may have a significant effect on liquid metal systems, e.g., tritium extraction systems. Liquid metals like LLE under nuclear irradiation in, e.g., breeding blankets of a nuclear fusion reactor would generate tritium which is to be extracted and recirculated as fuel. At the same time that tritium is bred, helium is also generated and may precipitate in the form of nano bubbles...

  4. Surface Nanocrystallization of 3Cr13 Stainless Steel Induced by High-Current Pulsed Electron Beam Irradiation

    Directory of Open Access Journals (Sweden)

    Zhiyong Han

    2013-01-01

    Full Text Available The nanocrystalline surface was produced on 3Cr13 martensite stainless steel surface using high-current pulsed electron beam (HCPEB technique. The structures of the nanocrystallized surface were characterized by X-ray diffraction and electron microscopy. Two nanostructures consisting of fine austenite grains (50–150 nm and very fine carbides precipitates are formed in melted surface layer after multiple bombardments via dissolution of carbides and crater eruption. It is demonstrated that the dissolution of the carbides and the formation of the supersaturated Fe (C solid solution play a determining role on the microstructure evolution. Additionally, the formation of fine austenite structure is closely related to the thermal stresses induced by the HCPEB irradiation. The effects of both high carbon content and high value of stresses increase the stability of the austenite, which leads to the complete suppression of martensitic transformation.

  5. Extreme embrittlement of austenitic stainless steel irradiated to 75-81 dpa at 335-360{degrees}C

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

    It is generally accepted that void swelling of austenitic steels ceases below some temperature in the range 340-360{degrees}C, and exhibits relatively low swelling rates up to {approximately}400{degrees}C. This perception may not be correct at all irradiation conditions, however, since it was largely developed from data obtained at relatively high displacement rates in fast reactors whose inlet temperatures were in the range 360-370{degrees}C. There is an expectation, however, that the swelling regime can shift to lower temperatures at low displacement rates via the well-known {open_quotes}temperature shift{close_quotes} phenomenon. It is also known that the swelling rates at the lower end of the swelling regime increase continuously at a sluggish rate, never approaching the terminal 1%/dpa level within the duration of previous experiments. This paper presents the results of an experiment conducted in the BN-350 fast reactor in Kazakhstan that involved the irradiation of argon-pressurized thin-walled tubes (0-200 MPa hoop stress) constructed from Fe-16Cr-15Ni-3Mo-Nb stabilized steel in contact with the sodium coolant, which enters the reactor at {approx}270{degrees}C. Tubes in the annealed condition reached 75 dpa at 335{degrees}C, and another set in the 20% cold-worked condition reached 81 dpa at 360{degrees}C. Upon disassembly all tubes, except those in the stress-free condition, were found to have failed in an extremely brittle fashion. The stress-free tubes exhibited diameter changes that imply swelling levels ranging from 9 to 16%. It is expected that stress-enhancement of swelling induced even larger swelling levels in the stressed tubes.

  6. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  7. Weldability of reduced activation ferritic/martensitic steel under ultra power density fiber laser welding

    Energy Technology Data Exchange (ETDEWEB)

    Serizawa, H.; Kawahito, Y.; Katayama, S. [Osaka Univ., Joining and Welding research Institute, Ibaraki (Japan); Ogiwara, H.; Tanigawa, H. [Japan Atomic Energy Agency, Ibaraki-ken (Japan)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems as it has been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. As one of RAFMS, F82H, which has been developed and studied in Japan, is designed with emphasis on high temperature property and weldablility, and was provided and evaluated in various countries as a part of the collaboration of IEA fusion materials development. Although F82H is the well perceived RAFM as ITER Test Blanket Module (TBM) structural material, the weldability was proved though TIG, EB and YAG laser weld tests using only 15 and 25 mm thickness plate. In order to reduce the welding distortion, the residual stress and the area of the heat affected zone, it is necessary to decrease the total heat input under the welding. Recently, as a result of R and D efforts about the sources of laser beam, a high-power fiber laser beam has been developed as one of the desirable heat sources for high-speed and deep-penetration welding. Since the power density of the fiber laser beam is very large, it is possible to increase the welding speed more than 10 m/min. So, in this study, the weldability of 1.5 mm thickness F82H plate and pipe was examined by using a ultra power density fiber laser, in order to reveal the excellent weldability of F82H. As a basic study of the butt welding between 1.5 mm plate and 1.5 mm thickness pipe with 11 mm outer diameter, the focus position, the beam position and the laser power were varied using 25 mm square plate and 25 mm length pipe. Then, by using the fiber laser with 1.1 MW/mm{sup 2} peak power density under the appropriate welding condition obtained from the basic study, a full penetrated weld bead with narrow width was formed in the butt

  8. Field emission study from an array of hierarchical micro protrusions on stainless steel surface generated by femtosecond pulsed laser irradiation

    Science.gov (United States)

    Singh, A. K.; Suryawanshi, Sachin R.; More, M. A.; Basu, S.; Sinha, Sucharita

    2017-02-01

    This paper reports our results on femtosecond (fs) pulsed laser induced surface micro/nano structuring of stainless steel 304 (SS 304) samples and their characterization in terms of surface morphology, formed material phases on laser irradiation and field emission studies. Our investigations reveal that nearly uniform and dense array of hierarchical micro-protrusions (density: ∼5.6 × 105 protrusions/cm2) is formed upon laser treatment. Typical tip diameters of the generated protrusions are in the range of 2-5 μm and these protrusions are covered with submicron sized features. Grazing incidence X-ray diffraction (GIXRD) analysis of the laser irradiated sample surface has shown formation mainly of iron oxides and cementite (Fe3C) phases in the treated region. These laser micro-structured samples have shown good field emission properties such as low turn on field (∼4.1 V/μm), high macroscopic field enhancement factor (1830) and stable field emission current under ultra high vacuum conditions.

  9. Views of TAGSI on effects of neutron irradiation on ductile tearing in ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Knott, J.F. [School of Metallurgy and Materials, University of Birmingham, Birmingham B15 2TT (United Kingdom); Lidbury, D.P.G. [Serco Technical and Assurance Services, Walton House, 404 Faraday Street, Birchwood Park, Warrington WA3 6GA (United Kingdom)], E-mail: david.lidbury@serco.com

    2009-07-15

    The paper reviews information pertaining to effects of neutron irradiation on 'upper-shelf' Charpy impact behaviour and on elastic/plastic fracture mechanics characterising parameters, again for 'upper shelf' conditions, in which the initiation and early growth of a crack involve ductile tearing. The hardening and associated reduction in strain-hardening capacity induced by irradiation gives rise to a decrease in Charpy upper shelf energy. Effects on J-based parameters are more complicated. The material resistance parameters tend to increase for low dose, but decrease at high dose, when the decrease in crack-tip ductility outweighs the effect of hardening. High doses can produce 'fast shear' fracture, which propagates rapidly and is therefore more likely to induce brittle cleavage fracture. The situation is exacerbated if the irradiation also promotes inter-granular segregation and fracture, hence reducing the local brittle fracture stress. For the levels of irradiation experienced by the types of UK civil reactors in operation, no fracture instability is expected to arise as a result of ductile fracture mechanisms alone.

  10. Cavity nucleation and growth in dual beam irradiated 316L industrial austenitic stainless steel

    Science.gov (United States)

    Jublot-Leclerc, S.; Li, X.; Legras, L.; Fortuna, F.; Gentils, A.

    2017-10-01

    Thin foils of 316L were simultaneously ion irradiated and He implanted in situ in a Transmission Electron Microscope at elevated temperatures. The resulting microstructure is carefully investigated in comparison with previous single ion irradiation experiments with a focus on the nucleation and growth of cavities. Helium is found to strongly enhance the nucleation of cavities in dual beam experiments. On the contrary, it does not induce more nucleation when implanted consecutively to an in situ ion irradiation but rather the growth of cavities by absorption at existing cavities, which shows the importance of synergistic effects and He injection mode on the microstructural changes. In both dual beam and single beam experiments, the characteristics of the populations of cavities, either stabilized by He or O atoms, are in qualitative agreement with the predictions of rate theory models for cavity growth. The evolutions of cavity population as a function of irradiation conditions can be reasonably well explained by the concept of relative sink strength of cavities and dislocations and the resulting partitioning of defects at sinks, or conversely recombination when either of the sinks dominates. The dislocations whose presence is a prerequisite to cavity growth in rate theory models are not observed in all studied conditions. In this case, the net influx of vacancies to cavities necessary to their growth and conversion to voids is believed to result from free surface effects, and possibly also segregation of elements close to the cavity surface. In any studied condition, the measured swelling is low, which is ascribed to the dilution of gaseous atoms among a high density of cavities as well as a high rate of point defect recombination and loss at traps. This high rate of recombination enhanced when dislocations are absent appears to result in the formation of overpressurized He bubbles.

  11. Electrochemical behaviour of gold and stainless steel under proton irradiation and active RedOx couples

    Energy Technology Data Exchange (ETDEWEB)

    Leoni, E. [Commissariat a l' Energie Atomique, DEN/DANS/DPC/SCCME, CEA-Saclay, 91191 Gif sur Yvette (France)], E-mail: elisa.leoni@polytechnique.edu; Corbel, C. [Laboratoire des Solides Irradies, Ecole Polytechnique, 91128 Palaiseau (France)], E-mail: catherine.corbel@polytechnique.fr; Cobut, V. [Laboratoire Atomes et Molecules en Astrophysique, Universite de Cergy-Pontoise, 5 Mail Gay-Lussac, Neuville/Oise, 95031 Cergy-Pontoise Cedex (France); Simon, D. [CNRS-CERI 3a rue de la Ferollerie, 45071 Cedex 2 Orleans (France); Feron, D. [Commissariat a l' Energie Atomique, DEN/DANS/DPC/SCCME, CEA-Saclay, 91191 Gif sur Yvette (France)], E-mail: Damien.FERON@cea.fr; Roy, M.; Raquet, O. [Commissariat a l' Energie Atomique, DEN/DANS/DPC/SCCME, CEA-Saclay, 91191 Gif sur Yvette (France)

    2007-12-01

    Model experiments are reported where proton beams delivered by the cyclotron located at CERI (CNRS-Orleans) are used for irradiating AISI 316L/water and Au/water high purity interfaces with 6 MeV protons. The free exchange potentials at the interfaces are recorded as a function of time at room temperature in situ before, under, and after proton irradiation. The evolutions are compared to those calculated for the Nernst potentials associated with the radiolytic RedOx couples. It is shown how the comparison gives evidence that five radiolytic species - O{sub 2}{center_dot}, H{sub 2}O{sub 2}, HO{sub 2}{sup -}, HO{sub 2}{center_dot} and O{sub 2}{center_dot}{sup -} exchange electrons at the Au interfaces in a range of dose rates that vary over three orders of magnitudes, i.e. 0.0048 < dr(10{sup 7} Gy h{sup -1}) < 4.8. The balance between the electron exchanges at Au interfaces is adjusted by the RedOx reactions associated with the above species. The free exchange potential reaches the same steady value for Au and AISI 316L interfaces irradiated at high doses, {>=}2.5 x 10{sup 7} Gy, (0.020 {+-} 0.025) V versus NHE. Such low values are the first ones to be reported. The HO{sub 2}{center_dot} and O{sub 2}{center_dot}{sup -} radical disproportionations play a key role and control the potential at the interfaces under 6 MeV proton flux. This role is generally mostly overlooked for gamma irradiation.

  12. Effect of activation cross section uncertainties in the assessment of primary damage for MFE/IFE low-activation steels irradiated in IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)], E-mail: cabellos@din.upm.es; Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, 28045 Madrid (Spain); Garcia-Herranz, N. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid (UPM), C/Jose Gutierrez Abascal, n2, 28006 Madrid (Spain); Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Otero, B. [Dept. de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain)

    2009-04-30

    The present study is mainly aimed to provide the primary damage (displacements per atom, generation of solid transmutants and gas production rates) of structural materials irradiated in the high and medium flux test modules of the International Fusion Materials Irradiation Facility (IFMIF). We have investigated if the change of the composition during the irradiation time has effect on the prediction of the atomic displacements. The effect of the activation cross section uncertainties in the assessment of both solid transmutants and hydrogen and helium production is also analyzed. The results are provided element-by-element, so that the primary damage of any material irradiated in such neutron environments can be easily assessed; in this paper, we have predicted the primary damage of the low activation steel Eurofer.

  13. A Hierarchical Upscaling Method for Predicting Strength of Materials under Thermal, Radiation and Mechanical loading - Irradiation Strengthening Mechanisms in Stainless Steels

    Energy Technology Data Exchange (ETDEWEB)

    Li, Dongsheng; Zbib, Hussein M.; Garmestani, Hamid; Sun, Xin; Khaleel, Mohammad A.

    2011-07-01

    Stainless steels based on Fe-Cr-Ni alloys are the most popular structural materials used in reactors. High energy particle irradiation of in this kind of polycrystalline structural materials usually produces irradiation hardening and embrittlement. The development of predictive capability for the influence of irradiation on mechanical behavior is very important in materials design for next-generation reactors. Irradiation hardening is related to structural information crossing different length scale, such as composition, dislocation, crystal orientation distribution and so on. To predict the effective hardening, the influence factors along different length scales should be considered. A multiscale approach was implemented in this work to predict irradiation hardening of iron based structural materials. Three length scales are involved in this multiscale model: nanometer, micrometer and millimeter. In the microscale, molecular dynamics (MD) was utilized to predict on the edge dislocation mobility in body centered cubic (bcc) Fe and its Ni and Cr alloys. On the mesoscale, dislocation dynamics (DD) models were used to predict the critical resolved shear stress from the evolution of local dislocation and defects. In the macroscale, a viscoplastic self-consistent (VPSC) model was applied to predict the irradiation hardening in samples with changes in texture. The effects of defect density and texture were investigated. Simulated evolution of yield strength with irradiation agrees well with the experimental data of irradiation strengthening of stainless steel 304L, 316L and T91. This multiscale model we developed in this project can provide a guidance tool in performance evaluation of structural materials for next-generation nuclear reactors. Combining with other tools developed in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the models developed will have more impact in improving the reliability of current reactors and affordability of new

  14. Molecular desorption of stainless steel vacuum chambers irradiated with 42 MeV/u lead ions

    CERN Document Server

    Mahner, E; Laurent, Jean Michel; Madsen, N

    2003-01-01

    In preparation for the heavy ion program of the Large Hadron Collider at CERN, accumulation and cooling tests with lead ion beams have been performed in the Low Energy Antiproton Ring. These tests have revealed that due to the unexpected large outgassing of the vacuum system, the dynamic pressure of the ring could not be maintained low enough to reach the required beam intensities. To determine the actions necessary to lower the dynamic pressure rise, an experimental program has been initiated for measuring the molecular desorption yields of stainless steel vacuum chambers by the impact of 4.2 MeV/u lead ions with the charge states +27 and +53. The test chambers were exposed either at grazing or at perpendicular incidence. Different surface treatments (glow discharges, nonevaporable getter coating) are reported in terms of the molecular desorption yields for H/sub 2 /, CH/sub 4/, CO, Ar, and CO/sub 2/. (16 refs).

  15. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  16. Microstructure of Au-ion irradiated 316L and FeNiCr austenitic stainless steels

    Science.gov (United States)

    Jublot-Leclerc, S.; Li, X.; Legras, L.; Lescoat, M.-L.; Fortuna, F.; Gentils, A.

    2016-11-01

    Thin foils of 316L were irradiated in situ in a Transmission Electron Microscope with 4 MeV Au ions at 450 °C and 550 °C. Similar irradiations were performed at 450 °C in FeNiCr. The void and dislocation microstructure of 316L is found to depend strongly on temperature. At 450 °C, a dense network of dislocation lines is observed in situ to grow from black dot defects by absorption of other black dots and interstitial clusters whilst no Frank loops are detected. At 550 °C, no such network is observed but large Frank loops and perfect loops whose sudden appearance is concomitant with a strong increase in void density as a result of a strong coupling between voids and dislocations. Moreover, differences in both alloys microstructure show the major role played by the minor constituents of 316L, increasing the stacking fault formation energy, and possibly leading to significant differences in swelling behaviour.

  17. Effect of neutron irradiation on the impact properties of A533B steel

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Kumar, A.S. [Univ. of Missouri, Rolla, MO (United States); Rosinski, S.T. [Sandia National Labs., Albuquerque, NM (United States); Hamilton, M.L. [Pacific Northwest Lab., Richland, WA (United States)

    1994-10-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material, ASTM type A 533 Grade B (A533B) having a low USE (USE < 100 J). The methodology appears to be more satisfactory than those methodologies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with specimen size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full, half, and third size specimens. The ductile-to-brittle transition temperature (DBTT) increased due to irradiation at 150 C to a nominal fluence of 1.0 {times} 10{sup 19} n/cm{sup 2} (E > 1 MeV) by 78 {degree}, 83{degree}, and 70{degree}C for full, half, and third size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry.

  18. Effect of specimen size on the impact properties of neutron irradiated A533B steel

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E. [Missouri Univ., Rolla, MO (United States). Dept. of Nuclear Engineering; Kumar, A.S. [Missouri Univ., Rolla, MO (United States). Dept. of Nuclear Engineering; Rosinski, S.T. [Sandia National Laboratories, MS-0741, Albuquerque, NM 87185 (United States); Hamilton, M.L. [Pacific Northwest Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

    1995-08-01

    A new methodology is proposed to correlate the upper shelf energy (USE) of full-size and subsize Charpy specimens of a nuclear reactor pressure vessel plate material. The methodology appears to be more satisfactory than those methodologies proposed earlier. The USE was normalized by a normalization factor involving the dimensions of the Charpy specimen, the elastic stress concentration factor, and the plastic constraint at the notch root. The normalized values of the USE were found to be invariant with specimen size. In addition, it was also found that the ratio of the USE of unirradiated to that of irradiated materials was approximately the same for full-, half-, and third-size specimens. The ductile-to-brittle transition temperature (DBTT) increased due to irradiation at 150 C to a nominal fluence of 1.0x10{sup 19} n/cm{sup 2} (E>1 MeV) by 78, 83, and 70 C for full-, half-, and third-size specimens, respectively. These shifts in DBTT appeared to be independent of specimen size and notch geometry. (orig.).

  19. Investigation on hot-dip aluminised and subsequent HIP'ped steel sheets

    Energy Technology Data Exchange (ETDEWEB)

    Glasbrenner, H.; Konys, J. E-mail: juergen.konys@imf.fzk.de

    2001-11-01

    Tritium permeation can be reduced significantly by a suitable coating on the structural material. Since alumina has the capability of tritium permeation reduction the development of such coatings on ferritic martensitic steels by hot-dip aluminising of F82H-mod. steel sheets was already performed successfully. An improvement of these coatings were achieved by subsequent HIP'ping at 1040 deg. C for 0.5 h at 250, 500 and 750 bar and subsequently tempered at 750 deg. C for 1 h at 1 bar. All samples were investigated by means of metallographical examination, EDX line scan analysis and Vickers micro hardness measurements. The high pressure produced two observed changes: firstly, with increasing pressure the thickness of the FeAl phase increases and the thickness of the {alpha}-Fe(Al) phase decreases, and secondly the formation of pores could be suppressed successfully. The Vickers micro hardness of the base material F82H-mod. is not influenced by the heat-treatment under pressure and is about 215 HV.

  20. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

    Science.gov (United States)

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-01

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  1. Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation.

    Science.gov (United States)

    Lu, Chenyang; Lu, Zheng; Wang, Xu; Xie, Rui; Li, Zhengyuan; Higgins, Michael; Liu, Chunming; Gao, Fei; Wang, Lumin

    2017-01-12

    The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

  2. Stainless steel wire mesh-supported ZnO for the catalytic photodegradation of methylene blue under ultraviolet irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Vu, Tan T.; Río, Laura del; Valdés-Solís, Teresa [Instituto Nacional del Carbón (INCAR-CSIC), c/Francisco Pintado Fe 26, 33011 Oviedo (Spain); Marbán, Gregorio, E-mail: greca@incar.csic.es [Instituto Nacional del Carbón (INCAR-CSIC), c/Francisco Pintado Fe 26, 33011 Oviedo (Spain)

    2013-02-15

    Highlights: ► Novel support of ZnO nanoparticles for photodegradation of organic contaminants. ► Small size of ZnO particles is key for high activity. ► The best catalyst has superior activity to TiO{sub 2} P25. ► Catalyst deactivation can be recognised by the reaction order. -- Abstract: The aim of this study was to assess the activity of catalysts formed by nanostructured zinc oxide supported on stainless steel wire mesh for the photocatalytic degradation of methylene blue under UV irradiation. Catalysts prepared by means of different low temperature synthesis methods, as described in a previous work (Vu et al., Mater. Res. Bull. 47 (2012) 1577–1586) were tested. A new activity parameter was introduced in order to compare the catalytic activity of the different catalysts. The best catalyst showed a catalytic activity higher than that of the reference material TiO{sub 2} P25 (Degussa-Evonik). This high activity is attributed to a higher quantum yield derived from the small particle length of the ZnO deposited on the wire mesh. The photocatalytic degradation kinetics of methylene blue fitted a potential model with n orders ranging from 0.5 to 6.9. Reaction orders over 1 were attributed to catalyst deactivation during the reaction resulting from the photocorrosion of ZnO.

  3. Charpy impact test results of four low activation ferritic alloys irradiated at 370{degrees}C to 15 DPA

    Energy Technology Data Exchange (ETDEWEB)

    Schubert, L.E.; Hamilton, M.L.; Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-10-01

    Miniature CVN specimens of four low activation ferritic alloys have been impact tested following irradiation at 370{degrees}C to 15 dpa. Comparison of the results with those of control specimens indicates that degradation in the impact behavior occurs in each of these four alloys. The 9Cr-2W alloy referred to as GA3X and the similar alloy F82H with 7.8Cr-2W appear most promising for further consideration as candidate structural materials in fusion energy system applications. These two alloys exhibit a small DBTT shift to higher temperatures but show increased absorbed energy on the upper shelf.

  4. Microstructure evolution of two model ferritic/martensitic steels under in situ ion irradiation at low doses (0–2 dpa)

    Energy Technology Data Exchange (ETDEWEB)

    Kaoumi, D., E-mail: djamelkaoumi@gmail.com [University of South Carolina, 300 Main St., Columbia, SC 29208 (United States); Adamson, J. [University of South Carolina, 300 Main St., Columbia, SC 29208 (United States); Kirk, M., E-mail: kirk@anl.gov [Argonne National Laboratory, Bldg 212, IL 60439 (United States)

    2014-02-01

    Ferritic/martensitic steels are candidate materials for structural and cladding components designed for Generation IV reactors because of their superior resistance to radiation damage at the high operating temperatures envisioned in these reactors. To enable the development and optimization of such advanced alloys for in-reactor use, a fundamental understanding of radiation damage accumulation in materials is required. In this work, two model F/M steels (12Cr model alloy and 9Cr model alloy) were irradiated with 1 MeV Kr ions at 50 K, 180 K, 298 K, 473 K and 573 K in situ in a TEM. The microstructure evolution under irradiation was followed and characterized at successive doses in terms of irradiation-induced defect formation and evolution, defect density, size distribution and interaction with the as-fabricated microstructure (e.g. dislocation networks, lath boundaries) using weak-beam dark-field imaging. The effect of the irradiation temperature on the defect kinetics is assessed at doses up to 2 dpa.

  5. A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, Ernest D. [Modeling and Computing Services, LLC; Odette, George Robert [UCSB; Nanstad, Randy K [ORNL; Yamamoto, Takuya [ORNL

    2007-11-01

    The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

  6. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Marini, B., E-mail: bernard.marini@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France); Averty, X. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SEMI (now DEN/DANS/DM2S/SEMT), F-91191 Gif-sur Yvette (France); Wident, P.; Forget, P.; Barcelo, F. [Commissariat à l' Energie Atomique et aux Energies Alternatives, DEN/DANS/DMN/SRMA, F-91191 Gif-sur Yvette (France)

    2015-10-15

    The hardening and the embrittlement under neutron irradiation of an A508 type RPV steel considering three different microstructures (bainite, bainite-martensite and martensite)have been investigated These microstructures were obtained by quenching after autenitization at 1100 °C. The irradiation induced hardening appears to depend on microstructure and is correlated to the yield stress before irradiation. The irradiation induced embrittlement shows a more complex dependence. Martensite bearing microstructures are more sensitive to non hardening embrittlement than pure bainite. This enhanced sensitivity is associated with the development of intergranular brittle facture after irradiation; the pure martensite being more affected than the bainite-martensite. It is of interest to note that this mixed microstructure appears to be more embrittled than the pure bainitic or martensitic phases in terms of temperature transition shift. This behaviour which could emerge from the synergy of the embrittlement mechanisms of the two phases needs further investigations. However, the role of microstructure on brittle intergranular fracture development appears to be qualitatively similar under neutron irradiation and thermal ageing.

  7. Comparison of laser welding conditions of Zircaloy-4 and stainless steel for nuclear fuel irradiation rig

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kahye; Hong, Jintae; Joung, Changyoung; Heo, Sungho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Various materials for Zircaloy-4, SUS 316L, such as Inconel, are used as a survey rig that has been produced for fuel irradiation testing. Precision sensors, thermocouples, LVDT, and SPND should also be assembled. Therefore, a welding device for connecting them is necessary. With a high density of energy, laser welding can be properly used in a deep permeation, and in precisely welding narrow and deep joints. In particular, it has been applied to other fields such as metal welding. Since the technology bears no pores or cavities, resulting in a clean surface after the welding process, it does not require an 'after-process' such as grinding or polishing, which is useful where high water-tightness is required. Therefore, we developed and researched a special fiber laser welding system for the production of a nuclear research rig. Through the above test, the different conditions of laser welding were found for Zircaloy-4 and AISI 316L used for producing a nuclear fuel research rig, performing the most optimal welding conditions according to the properties of the materials in the future.

  8. Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Wenhui [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Chen, Jihong; Luo, Fengfeng; Li, Tiecheng [Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education, Hubei Nuclear Solid Physics Key Laboratory and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, Wuhan University, Wuhan 430072 (China); Suo, Jinping; Yang, Feng [State Key Laboratory of Mould Technology, Institute of Materials Science and Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-04-15

    Highlights: • Helium/hydrogen synergistic effect can increase irradiation swelling of RAFM steel. • Hydrogen can be trapped to the outer surface of helium bubbles. • Too large a helium bubble can become movable. • Point defects would become mobile and annihilate at dislocations at high temperature. • The peak swelling temperature for RAFM steel is 450 °C. - Abstract: In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He{sup +} beam and sequential He{sup +} and H{sup +} beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.

  9. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends.

  10. The strong influence of displacement rate on void swelling in variants of Fe-16Cr-15Ni-3Mo austenitic stainless steel irradiated in BN-350 and BOR-60

    Energy Technology Data Exchange (ETDEWEB)

    Budylkin, N.I.; Bulanova, T.M.; Mironova, E.G.; Mitrofanova, N.M.; Porollo, S.I.; Chernov, V.M.; Shamardin, V.K.; Garner, F.A. E-mail: frank.garner@pnl.gov

    2004-08-01

    Recent irradiation experiments conducted on a variety of austenitic stainless steels have shown that void swelling appears to be increased when the dpa rate is decreased, primarily by a shortening of the transient regime of swelling. This paper presents results derived from nominally similar irradiations conducted on six Russian steels, all laboratory heat variants of Fe-16Cr-15Ni-3Mo-Nb-B, with each irradiated in two fast reactors, BOR-60 and BN-350. The BN-350 irradiation proceeded at a dpa rate three times higher than that conducted in BOR-60. In all six steels, a significantly higher swelling level was attained in BOR-60, agreeing with the results of earlier studies.

  11. Effect of gadolinium nitrate concentration on molecular product yield during gamma irradiation and on corrosion of stainless steel

    Science.gov (United States)

    Mal, D.; Puspalata, R.; Rangarajan, S.; Velmurugan, S.

    2017-09-01

    Effect of high concentrations of soluble neutron poison gadolinium nitrate, Gd(NO3)3, in the moderator system of a proposed advanced Indian nuclear reactor, was evaluated from the safety point of view. The radiolytic yields of H2 and H2O2 was expected to be high as moderator water system pH would be lowered and conductivity also would be high by the addition of higher concentration Gd(NO3)3 solutions during various shutdown states. Experiments were carried out to estimate this increase in radiolytic yield of molecular products with the addition of Gd(NO3)3 in the concentration range of 15-400 mg kg-1. Both the H2O2 and H2 yields were found to increase with absorbed dose and also with increasing Gd3+ concentration up to 100 mg kg-1 but the increase were marginal in 100-400 mg kg-1 range. For a given concentration of Gd(NO3)3 solution, radiolysis in high purity D2O showed a lower D2 formation than H2 in light water. In a simulated moderator temperature of 65 °C, a higher yield of H2 was observed. The headspace provided above the liquid phase in irradiation zone had shown to have a substantial effect on the generation of H2. With decreasing headspace, H2 generation increased and went through a maximum. Considering the expected long operational life ( 100 years) for the proposed reactor, the corrosion rate of the structural materials (stainless steel 304 LN) in contact with this high concentration Gd(NO3)3 solution was also estimated at 65 °C which showed a negligible effect.

  12. Correlation of microstructure with hardness and wear resistance in (TiC, SiC)/stainless steel surface composites fabricated by high-energy electron-beam irradiation

    Science.gov (United States)

    Yun, Eunsub; Kim, Yong Chan; Lee, Sunghak; Kim, Nack J.

    2004-03-01

    Stainless-steel-based surface composites reinforced with TiC and SiC carbides were fabricated by high-energy electron beam irradiation. Four types of powder/flux mixtures, i.e., TiC, (Ti + C), SiC, and (Ti + SiC) powders with 40 wt. pct of CaF2 flux, were deposited evenly on an AISI 304 stainless steel substrate, which was then irradiated with an electron beam. TiC agglomerates and pores were found in the surface composite layer fabricated with TiC powders because of insufficient melting of TiC powders. In the composite layer fabricated with Ti and C powders having lower melting points than TiC powders, a number of primary TiC carbides were precipitated while very few TiC agglomerates or pores were formed. This indicated that more effective TiC precipitation was obtained from the melting of Ti and C powders than of TiC powders. A large amount of precipitates such as TiC and Cr7C3 improved the hardness, high-temperature hardness, and wear resistance of the surface composite layer two to three times greater than that of the stainless steel substrate. In particular, the surface composite fabricated with SiC powders had the highest volume fraction of Cr7C3 distributed along solidification cell boundaries, and thus showed the best hardness, high-temperature hardness, and wear resistance.

  13. Post irradiation fatigue tests of type 316 LN stainless steel. Final report for the ITER Task T511, Subtask 1. European Technology Programme Task GB5-T217

    Energy Technology Data Exchange (ETDEWEB)

    Norring, K.; Koenig, M

    2002-01-01

    The main objective of this Subtask was to estimate the corrosion fatigue behaviour of 316L Stainless Steel (SS) and SS/SS joints, and to check among others the influence of irradiation. Joints were produced by solid Hot Isostatic Pressure (HIP) and powder HIP. Conventional material was used for comparison. The specimens were supplied by EFDA and were irradiated to 4 dpa in Dimitrovgrad (Russia). All specimens were tested at 150 deg C in hydrogenated high purity water. Testing was performed with a stepwise decrease in {delta}K keeping K{sub max} constant. The crack growth rates of irradiated as well as unirradiated specimens tested earlier are of the same magnitude, around 2x10{sup -5} mm/cycle at {delta}K= 18 MPa{radical}m. Thus, irradiation does not seem to enhance the fatigue crack growth rate, at least not up to irradiation levels of 4 dpa. But it is worth noting that the exponents in the da/dN versus {delta}K equation, also known as Paris' law, seems to fall within two areas, either around 3.5 or just below 2. Both Powder HIPed and Solid HIPed specimens are found in both groups. The reason for this is not evident. The fracture surfaces of the specimens show typical fatigue appearance.

  14. Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique

    Energy Technology Data Exchange (ETDEWEB)

    Otsuka, Teppei, E-mail: t-otsuka@nucl.kyushu-u.ac.jp [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan); Tanabe, Tetsuo [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Hakozaki 6-10-1, Higashi-ku, Fukuoka 812-8581 (Japan)

    2013-12-15

    Highlights: ► We applied a tritium imaging plate technique to depth profiling of hydrogen in bulk. ► Changes of hydrogen depth profiles in the steel by thermal annealing were examined. ► We proposed a release model of plasma-loaded hydrogen in the steel. ► Hydrogen is trapped at trapping sites newly developed by plasma loading. ► Hydrogen is also trapped at surface oxides and hardly desorbed by thermal annealing. -- Abstract: In order to understand how hydrogen loaded by plasma in F82H is removed by annealing at elevated temperatures in vacuum, depth profiles of plasma-loaded hydrogen were examined by means of a tritium imaging plate technique. Owing to large hydrogen diffusion coefficients in F82H, the plasma-loaded hydrogen easily penetrates into a deeper region becoming solute hydrogen and desorbs by thermal annealing in vacuum. However the plasma-loading creates new hydrogen trapping sites having larger trapping energy than that for the intrinsic sites beyond the projected range of the loaded hydrogen. Some surface oxides also trap an appreciable amount of hydrogen which is more difficult to remove by the thermal annealing.

  15. Observation and rate theory modeling of grain boundary segregation in Σ3 twin boundaries in ion-irradiated stainless steel 316

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Gyeong-Geun, E-mail: gglee@kaeri.re.kr; Jin, Hyung-Ha; Lee, Yong-Bok; Kwon, Junhyun

    2014-06-01

    Radiation-induced segregation (RIS) is the phenomenon of compositional change at point defect sinks in alloys irradiated at a moderate temperature. Owing to the potential relevance of RIS by way of the susceptibility of structural materials to irradiation-assisted stress corrosion cracking, basic research on austenitic stainless steels used in nuclear reactors has been carried out in recent years. In this work, commercial stainless steel 316 specimens were irradiated with Fe ions, and the resulting changes in Cr and Ni compositions were characterized using transmission electron microscopy and energy-dispersive X-ray spectroscopy. The samples with various grain boundary orientations, including the special Σ3 orientation, were analyzed. The ledges of a few special Σ3 twin boundaries showed significantly higher RIS compared to the coherent regions. The RIS behavior of a parallel twin pair was observed, and two profiles of RIS were found in them. The inner twins in multi-twins showed considerably lower RIS compared to the outer twins. For the calculation of RIS, time-dependent differential equations based on the rate theory were established and numerically integrated. An additional variable, representing the sink strength of the grain boundary, was introduced in the differential equations, and the concentration profiles of the Σ3 twins were calculated. The calculated results were in good agreement with the experimental results.

  16. Dependence of mode I and mixed mode I/III fracture toughness on temperature for a ferritic/martensitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, H.; Jones, R.H.; Gelles, D.S. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1995-04-01

    The objective is to investigate the dependence of mode I and mixed mode I/III fracture toughness on temperature in the range of {minus}95{degrees}C to 25{degrees}C for a low activation ferritic/martensitic stainless steel (F82-H). Mode I and mixed Mode I/III fracture toughnesses were investigated in the range of {minus}95 to 25{degree}C for a F82-H steel heat-treated in the following way; 1000{degree}C/20 h/air-cooled (AC), 1100{degree}C/7 min/AC, and 700{degree}C/2 h/AC. The results indicate that crack tip plasticity was increased by mixed mode loading, and suggest that at low temperature, mode I fracture toughness is the critical design parameter, but at temperatures above room temperature, expecially concerning fatigure and creep-fatigue crack growth rate, a mixed mode loading may be more harmful than a mode I loading for this steel because a mixed mode loading results in lower fracture toughness and higher crack tip plasticity (or dislocation activity).

  17. Modeling precipitation thermodynamics and kinetics in type 316 austenitic stainless steels with varying composition as an initial step toward predicting phase stability during irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Jae-Hyeok, E-mail: jhshim@kist.re.kr [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); High Temperature Energy Materials Research Center, Korea Institute of Science and Technology, Seoul 136-791 (Korea, Republic of); Povoden-Karadeniz, Erwin [Christian Doppler Laboratory for Early Stages of Precipitation, Vienna University of Technology, A-1040 Vienna (Austria); Kozeschnik, Ernst [Institute of Materials Science and Technology, Vienna University of Technology, A-1040 Vienna (Austria); Wirth, Brian D. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States)

    2015-07-15

    Highlights: • We model the precipitation kinetics in irradiated 316 austenitic stainless steels. • Radiation-induced phases are predicted to form at over 10 dpa segregation conditions. • The Si content is the most critical for the formation of radiation-induced phases. - Abstract: The long-term evolution of precipitates in type 316 austenitic stainless steels at 400 °C has been simulated using a numerical model based on classical nucleation theory and the thermodynamic extremum principle. Particular attention has been paid to the precipitation of radiation-induced phases such as γ′ and G phases. In addition to the original compositions, the compositions for radiation-induced segregation at a dose level of 5, 10 or 20 dpa have been used in the simulation. In a 316 austenitic stainless steel, γ′ appears as the main precipitate with a small amount of G phase forming at 10 and 20 dpa. On the other hand, G phase becomes relatively dominant over γ′ at the same dose levels in a Ti-stabilized 316 austenitic stainless steel, which tends to suppress the formation of γ′. Among the segregated alloying elements, the concentration of Si seems to be the most critical for the formation of radiation-induced phases. An increase in dislocation density as well as increased diffusivity of Mn and Si significantly enhances the precipitation kinetics of the radiation-induced phases within this model.

  18. Analysis of nano-sized irradiation-induced defects in Fe-base materials by means of small angle neutron scattering and molecular dynamics simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yu, G.

    2008-12-15

    of RAFM steels with Molecular Dynamics (MD) simulations of main expected nano-sized defects in irradiated pure Fe and Fe-He alloys, as model materials for RAFM steels, and simulations of their corresponding TEM images and SANS signals. In particular, the SANS signal of various types of defects was simulated for the first time. The methodology used in this work was the following: (i) SANS experiments were performed by applying a strong saturating magnetic field to unirradiated and irradiated specimens of three types of RAFM steels, namely the European EUROFER 97, the Japanese F82H and the Swiss OPTIMAX A steels. The available irradiated specimens included specimens which had been irradiated with 590 MeV protons in the Proton IRradiation EXperiment (PIREX) facility at the Paul Scherrer Institute (PSI) at temperatures in the range of 50-350 °C to doses in the range of 0.3-2.0 dpa. SANS spectra as well as values of the so-called A ratio, which represents the ratio of the total scattered intensity to the nuclear scattered intensity, were obtained for the various irradiation doses and temperatures investigated. (ii) MD simulations of atomic displacement cascades in pure Fe and in Fe-He alloys were performed using Embedded Atom Method (EAM) many-body interatomic potentials. The main nano-sized defects that should be produced in RAFM steels under irradiation were created by means of MD in pure Fe. These included dislocation loops of various types, voids, helium bubbles with various He concentration and Cr precipitates. (iii) TEM images of cascade damage and all the defects created by MD were simulated in the dark field/weak beam imaging modes by using the Electron Microscopy Software (EMS) developed by P.A. Stadelmann (EPFL) and analyzed in terms of variations of contrast intensities versus depth inside the specimen. (iv) The SANS signal provided by cascade damage and all the defects created by MD was simulated by using a slightly modified version of EMS, accounting for

  19. Long-term properties of reduced activation ferritic/martensitic steels for fusion reactor blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Shiba, Kiyoyuki, E-mail: Shiba.kiyoyuki@jaea.go.jp [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan); Tanigawa, Hiroyasu; Hirose, Takanori; Sakasegawa, Hideo; Jitsukawa, Shiro [Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2011-12-15

    Thermal aging properties of reduced activation ferritic/martensitic steel F82H was researched. The aging was performed at temperature ranging from 400 Degree-Sign C to 650 Degree-Sign C up to 100,000 h. Microstructure, precipitates, tensile properties, and Charpy impact properties were carried out on aged materials. Laves phase was found at temperatures between 550 and 650 Degree-Sign C and M{sub 6}C type carbides were found at the temperatures between 500 and 600 Degree-Sign C over 10,000 h. These precipitates caused degradation in toughness, especially at temperatures ranging from 550 Degree-Sign C to 650 Degree-Sign C. Tensile properties do not have serious aging effect, except for 650 Degree-Sign C, which caused large softening even after 10,000 h. Increase of precipitates also causes some degradation in ductility, but it is not critical. Large increase in ductile-to-brittle transition temperature was observed in the 650 Degree-Sign C aging. It was caused by the large Laves phase precipitation at grain boundary. Laves precipitates at grain boundary also degrades the upper-shelf energy of the aged materials. These aging test results indicate F82H can be used up to 30,000 h at 550 Degree-Sign C.

  20. Radiation-induced segregation and corrosion behavior on Σ3 coincidence site lattice and random grain boundaries in proton-irradiated type-316L austenitic stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Sakaguchi, N., E-mail: sakaguchi@eng.hokudai.ac.jp [Center for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Hokkaido (Japan); Endo, M.; Watanabe, S. [Center for Advanced Research of Energy and Materials, Faculty of Engineering, Hokkaido University, Sapporo 060-8628, Hokkaido (Japan); Kinoshita, H. [Fukushima National College of Technology, Iwaki 970-8034, Fukushima (Japan); Yamashita, S. [Fuels and Materials Department, O-arai Research and Development Center, Japan Atomic Energy Agency, Ibaraki 311-1393 (Japan); Kokawa, H. [Graduate School of Engineering, Tohoku University, Sendai 980-8579 (Japan)

    2013-03-15

    The behavior of radiation-induced segregation (RIS) and intergranular corrosion at random grain boundaries and Σ3 coincidence site lattice (CSL) boundaries in proton-irradiated 316L stainless steel was examined. The frequency of the CSL boundaries was enhanced up to 86.6% by grain boundary engineering treatment prior to irradiation. Significant nickel enrichment and chromium depletion were induced at the random grain boundary owing to the RIS. At faceted Σ3 CSL boundaries, chromium depletion occurred at the asymmetrical boundary facet plane whereas no RIS was observed at the coherent twin boundary. After the electrochemical etching test, an intergranular corrosion groove was found along the random grain boundaries because of the low chromium concentration (∼12%) at the boundaries. At the faceted Σ3 CSL boundaries, the discontinuous groove along the asymmetric facet plane was completely disrupted by the non-corrosive coherent twin boundary.

  1. Evaluation of the cross-sections of threshold reactions leading to the production of long-lived radionuclides during irradiation of steels by thermonuclear spectrum neutrons

    CERN Document Server

    Blokhin, A I; Manokhin, V N; Mikhajlyukova, M V; Nasyrova, S M; Skripova, M V

    2001-01-01

    The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,alpha) and (n,n alpha) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides sup 3 sup 9 Ar, sup 4 sup 2 Ar, sup 4 sup 1 Ca, sup 5 sup 3 Mn, sup 6 sup 0 Fe, sup 6 sup 0 Co, sup 5 sup 9 Ni and sup 6 sup 3 Ni. A comparison is made with the experimental and other evaluated data.

  2. Effect of chloride ion on corrosion behavior of SUS316L-grade stainless steel in nitric acid solutions containing seawater components under γ-ray irradiation

    Science.gov (United States)

    Sano, Y.; Ambai, H.; Takeuchi, M.; Iijima, S.; Uchida, N.

    2017-09-01

    Concerning the Fukushima Daiichi nuclear power plant accident, we investigated the effect of chloride ion on the corrosion behavior of SUS316L stainless steel, which is a typical material for the equipment used in reprocessing, in HNO3 solution containing seawater components, including under γ-ray irradiation condition. Electrochemical and immersion tests were carried out using a mixture of HNO3 and artificial seawater (ASW). In the HNO3 solution containing high amounts of ASW, the cathodic current densities increased and uniform corrosion progressed. This might be caused by strong oxidants, such as Cl2 and NOCl, generated in the reaction between HNO3 and Cl- ions. The corrosion rate decreased with the immersion time at low concentrations of HNO3, while it increased at high concentrations. Under γ-ray irradiation condition, the corrosion rate decreased due to the suppression of the cathodic reactions by the reaction between the above oxidants and HNO2 generated by radiolysis.

  3. Gas bubbles evolution peculiarities in ferritic-martensitic and austenitic steels and alloys under helium-ion irradiation

    NARCIS (Netherlands)

    Chernov, [No Value; Kalashnikov, AN; Kahn, BA; Binyukova, SY

    2003-01-01

    Transmission electron microscopy has been used to investigate the gas bubble evolution in model alloys of the Fe C system, ferritic-martensitic steels of 13Cr type, nickel and austenitic steels under 40-keV helium-ion it. radiation up to a fluence of 5 x 10(20) m(-2) at the temperature of 920 K. It

  4. Elastic modules and thermal conductivity of neutron irradiated type 13Cr2MoNbVB ferritic—martensitic steel

    Science.gov (United States)

    Zakharova, M. I.; Artemov, N. A.; Petrov, D. V.

    1996-10-01

    The temperature dependencies of elastic modulus, internal friction, Poisson ratio in the range from 20 to 600°C as well as of thermal conductivity and electrical resistance at temperatures from 20 to 1000°C have been determined for type 13Cr2MoNbVB ferritic—martensitic steel irradiated at 280°C in the BN-350 fast reactor to neutron fluence of 4.03 × 10 26 n/m 2 ( E > 0.1 MeV). During isochronal annealings at temperatures up to 0.65 Tm the recovery of the properties for irradiated steel has been investigated in the range of homologous temperatures from 0.28 Tm to 0.53 Tm. A pronounce recovery was observed at four substages in the temperature interval from 380 to 620°C. The activation energies determined for all recovery substages varied from 1.11 eV to 4.09 eV.

  5. Heat treatment effects on impact toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 100 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Alexander, D.J. [Oak Ridge National Lab., TN (United States)

    1997-08-01

    Plates of 9Cr-1MoVNb and 12Cr-1MoVW steels were given four different heat treatments: two normalizing treatments were used and for each normalizing treatment two tempers were used. Miniature Charpy specimens from each heat treatment were irradiated to {approx}19.5 dpa at 365{degrees}C and to {approx}100 dpa at 420{degrees}C in the Fast Flux Test Facility (FFTF). In previous work, the same materials were irradiated to 4-5 dpa at 365{degrees}C and 35-36 dpa at 420{degrees}C in FFTF. The tests indicated that prior austenite grain size, which was varied by the different normalizing treatments, had a significant effect on impact behavior of the 9Cr-1MoVNb but not on the 12Cr-1MoVW. Tempering treatment had relatively little effect on the shift in DBTT for both steels. Conclusions are presented on how heat treatment can be used to optimize impact properties.

  6. Post-irradiation annealing of Ni–Mn–Si-enriched clusters in a neutron-irradiated RPV steel weld using Atom Probe Tomography

    Energy Technology Data Exchange (ETDEWEB)

    Styman, P.D., E-mail: paul.styman@materials.ox.ac.uk [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Hyde, J.M. [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Parfitt, D.; Wilford, K. [Rolls-Royce, PO BOX 2000, Raynesway, Derby DE21 7XX (United Kingdom); Burke, M.G. [School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); English, C.A. [National Nuclear Laboratory, 168 Harwell Business Centre, Didcot, Oxon OX11 0QT (United Kingdom); Department of Materials, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); School of Materials, University of Manchester, Manchester M13 9PL (United Kingdom); Efsing, P. [Vattenfall Ringhals AB, Väröbacka (Sweden)

    2015-04-15

    Highlights: • Characterisation of high Ni neutron irradiated RPV surveillance samples at high fluence. • Post-irradiation annealing performed to give insight into the formation mechanisms of Ni–Mn–Si precipitates. • Dissolution of Ni–Mn–Si clusters appears to be lead by the removal of Mn. - Abstract: Atom Probe Tomography has been performed on as-irradiated and post-irradiation annealed surveillance weld samples from Ringhals Unit 3. The weld contains low Cu (0.07 at.%) and high Ni (1.5 at.%). A high number density (∼4 × 10{sup 23} m{sup −3}) of Ni–Mn–Si-enriched clusters was observed in the as-irradiated material. The onset of recovery was observed during the annealing for 30 min at 450 °C. Much more significant dissolution of clusters occurred during the 10 min 500 °C anneal, resulting in a reduction in mean cluster size and a halving of their volume fraction. Detailed analyses of the changes in microstructure demonstrate that the dissolution process is driven by migration of Mn atoms from the clusters. This may indicate a strong correlation between Mn and point defects. Dissolution of the clusters is shown to correlate with recovery of mechanical properties in this material.

  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

    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.

  8. Mechanical Property of China A508-3 Steel after Neutron Irradiation%国产 A508-3钢辐照性能

    Institute of Scientific and Technical Information of China (English)

    林赟; 宁广胜; 张长义; 佟振峰; 杨文

    2016-01-01

    Reactor pressure vessel (RPV ) is the critical un‐changeable component of the reactor during its service lifetime , w hich prevents the radioactive leak of the nuclear power plant core .The irradiation test (about 10 × 1019 cm -2 ,E≥1 MeV) of the pres‐sure vessel material of China A508‐3 steel in research reactor was carried out ,and the mechanics performance tests were carried out after the neutron irradiation ,including tensile property and impact property .The results show that the yield strength increases by 83 ,108 and 52 MPa ,and the tensile strength increases by 58 ,61 and 49 MPa at-100 ,20 and 288 ℃ , respectively . The ductile‐brittle transition temperature T41J increases by 68 ℃ ,and the upper shelf energy decreases by 61 J .Meanwhile ,by compa‐ring the property of un‐irradiated and irradiated material ,after irradiated to the level of 60 a service life ,A508‐3 steel still meets the reactor operation requirement .%反应堆压力容器(RPV )作为反应堆寿期内不可更换的核心设备,是防止堆芯放射性泄漏的最主要屏障。本文针对国产压力容器材料A508‐3钢,开展了一定剂量水平(约10×1019 cm-2,E≥1 M eV )的研究堆加速辐照试验,并进行了辐照后力学性能测试分析,包括拉伸性能和冲击性能测试。结果显示,辐照后在-100、20、288℃下,A508‐3钢的屈服强度分别增加了83、108、52 M Pa ,抗拉强度分别增加了58、61、49 M Pa ,韧脆转变温度 T41J增加了68℃,上平台能量降低了61 J。A508‐3钢辐照前后性能测试结果表明,在中子辐照至60 a寿期后,A508‐3钢仍能满足反应堆使用要求。

  9. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  10. Stabilization of the spark-discharge point on a sample surface by laser irradiation for steel analysis.

    Science.gov (United States)

    Matsuta, Hideyuki; Kitagawa, Kuniyuki; Wagatsuma, Kazuaki

    2006-10-01

    A combined technique with laser irradiation is suggested to control spark discharge for analytical use, having a unique feature that firing points of the spark discharge can be fixed by laser irradiation. Because the spark discharge easily initiates at particular surface sites, such as non-metallic inclusions, called selective discharge, the concentration of some elements sometimes deviates from their average one in spark discharge optical emission spectrometry. Therefore, stabilization of firing points on a sample surface could improve the analytical precision.

  11. Heavy-Section Steel Irradiation Program. Volume 2, No. 2: Semiannual progress report, April--September 1991

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1994-10-01

    Goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel stools as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and post-irradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is into 10 tasks: (1) program management, (2) K{sub Ic} curve shift in high-copper welds, (3) K{sub Ia} curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub Ic} and K{sub Ia} curve shifts in low upper-shelf welds, (6) irradiation effects in a commercial low upper-sheer weld, (7) microstructural analysis of irradiation effects, (8) in-service aged material evaluations, (9) correlation monitor materials, and (10) special technical assistance. This report provides an overview of the activities within each of these tasks from April to September 1991.

  12. Experimental studies of the effect of irradiation on the anaerobic corrosion of carbon steel in relation to the Belgian supercontainer concept

    Directory of Open Access Journals (Sweden)

    Reddy B.

    2011-04-01

    Full Text Available This paper describes recent results from an investigation of the effects of γ-radiation on the anaerobic corrosion of carbon steel in cement, in relation to the Belgian Supercontainer Concept for radioactive waste disposal. Anaerobic corrosion rates were measured by monitoring hydrogen evolution and the corresponding electrochemical behaviour was investigated by measuring open circuit potential and linear polarisation resistance. The test medium was alkaline simulated porewater, at γ-irradiation dose rates of 0 and 25 Gy hr−1, temperatures of 25 °C and 80 °C and chloride concentrations of 0 and 100 mg/l. The effects of radiation on the corrosion behaviour were found to be small.

  13. Microstructural stability of ODS Fe–14Cr (–2W–0.3Ti steels after simultaneous triple irradiation

    Directory of Open Access Journals (Sweden)

    M. Šćepanović

    2016-12-01

    Full Text Available Simultaneous triple-ion beam irradiation experiments with Fe5+, He+ and H+ ions were performed to simulate fusion damage on two nanostructured ferritic alloys with nominal composition Fe–14Cr–0.3Y2O3 and Fe–14Cr–2W–0.3Ti–0.3Y2O3. Samples were irradiated at 600°C to an estimated dose of ∼ 30dpa, 600appm He, 1500appm H, and the effects on the microstructure of these alloys investigated by analytical transmission electron microscopy. The results reveal the development of nanovoids, or small bubbles, undetected in the unirradiated samples, and a virtual compositional stability of the dispersion. Nevertheless, upon irradiation the measured size distribution indicates a slight growth of those dispersoids having the smaller sizes.

  14. Determination of elastic modulus and residual stress of plasma-sprayed tungsten coating on steel substrate

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmann Street 2, 85748 Garching (Germany)]. E-mail: jeong-ha.you@ipp.mpg.de; Hoeschen, T. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmann Street 2, 85748 Garching (Germany); Lindig, S. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmann Street 2, 85748 Garching (Germany)

    2006-01-01

    Plasma-sprayed tungsten, which is a candidate material for the first wall armour, shows a porous, heterogeneous microstructure. Due to its characteristic morphology, the properties are significantly different from those of its dense bulk material. Measurements of the elastic modulus of this coating have not been reported in the literature. In this work Young's modulus of highly porous plasma-sprayed tungsten coatings deposited on steel (F82H) substrates was measured. For the fabrication of the coating system the vacuum plasma-spray process was applied. Measurements were performed by means of three-point and four-point bending tests. The obtained modulus values ranged from 53 to 57 GPa. These values could be confirmed by the test result of a detached coating strip, which was 54 GPa. The applied methods produced consistent results regardless of testing configurations and specimen sizes. The errors were less than 1%. Residual stress of the coating was also estimated.

  15. Determination of elastic modulus and residual stress of plasma-sprayed tungsten coating on steel substrate

    Science.gov (United States)

    You, J. H.; Höschen, T.; Lindig, S.

    2006-01-01

    Plasma-sprayed tungsten, which is a candidate material for the first wall armour, shows a porous, heterogeneous microstructure. Due to its characteristic morphology, the properties are significantly different from those of its dense bulk material. Measurements of the elastic modulus of this coating have not been reported in the literature. In this work Young's modulus of highly porous plasma-sprayed tungsten coatings deposited on steel (F82H) substrates was measured. For the fabrication of the coating system the vacuum plasma-spray process was applied. Measurements were performed by means of three-point and four-point bending tests. The obtained modulus values ranged from 53 to 57 GPa. These values could be confirmed by the test result of a detached coating strip, which was 54 GPa. The applied methods produced consistent results regardless of testing configurations and specimen sizes. The errors were less than 1%. Residual stress of the coating was also estimated.

  16. Multicomponent Gas-Particle Flow and Heat/Mass Transfer Induced by a Localized Laser Irradiation on a Urethane-Coated Stainless Steel Substrate

    CERN Document Server

    Afrin, Nazia; Zhang, Yuwen; Chen, J K; Ritter, Robin; Lampson, Alan; Stohs, Jonathan

    2016-01-01

    A three-dimensional numerical simulation is conducted for a complex process in a laser-material system, which involves heat and mass transfer in a compressible gaseous phase and chemical reaction during laser irradiation on a urethane paint coated on a stainless steel substrate. A finite volume method (FVM) with a co-located grid mesh that discretizes the entire computational domain is employed to simulate the heating process. The results show that when the top surface of the paint reaches a threshold temperature of 560 K, the polyurethane starts to decompose through chemical reaction. As a result, combustion products CO2, H2O and NO2 are produced and chromium (III) oxide, which serves as pigment in the paint, is ejected as solid parcels from the paint into the gaseous domain. Variations of temperature, density and velocity at the center of the laser irradiation spot, and the concentrations of reaction reactant/products in the gaseous phase are presented and discussed, by comparing six scenarios with differen...

  17. Surface Nanocrystallization of 3Cr13 Stainless Steel Induced by High-Current Pulsed Electron Beam Irradiation

    OpenAIRE

    2013-01-01

    The nanocrystalline surface was produced on 3Cr13 martensite stainless steel surface using high-current pulsed electron beam (HCPEB) technique. The structures of the nanocrystallized surface were characterized by X-ray diffraction and electron microscopy. Two nanostructures consisting of fine austenite grains (50–150 nm) and very fine carbides precipitates are formed in melted surface layer after multiple bombardments via dissolution of carbides and crater eruption. It is demonstrated that th...

  18. Effects of impurity elements on mechanical properties and microstructures of reduced-activation ferritic/martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Sawahata, A. [Ibaraki Univ., Graduate School of Science and Engineering, Hitachi (Japan); Tanigawa, H.; Shiba, K. [Japan Atomic Energy Agency, Naga-gun, Ibaraki-ken (Japan); Enomoto, M. [Ibaraki Univ., Dept. of Materials Science, Faculty of Engineering, Hitachi (Japan)

    2007-07-01

    Full text of publication follows: Reduced activation ferritic/martensitic steels (RAFs), such as F82H (Fe-8Cr-2W-0.2V- 0.04Ta-0.1C, in wt%), are one of the leading candidates for structural materials of fusion reactors. Impact property of F82H can be improved by adjusting the amount of tantalum or titanium concentration. On the other hand, it was reported by microstructure analyses of IEA steel that tantalum has a tendency to form oxides and causes a large dispersion of fracture toughness. In this study, the correlation between titanium or tantalum concentration and the impact property were reported focusing on difference in microstructure. Charpy impact test and microstructure analyses were carried out against modified F82H series of which titanium, nitrogen and tantalum composition were controlled. Charpy impact test results showed that the ductile-brittle transition temperature (DBTT) of T05A (0.05Ta- 0.0014N-steels. The size distribution analyses of oxides indicate that the number density of composite oxides in T05B was higher than in T05A. In addition, EDX analyses showed that composite oxides in T05B had a strong peak of titanium, but the peak were not detected in the oxides in T05A. These results suggest that titanium has a significant influence on the formation of oxides, and affects the impact property. The influence of tantalum concentration on the formation of these oxides and mechanical properties will be reported. (authors)

  19. Microstructural behavior of VVER-440 reactor pressure vessel steels under irradiation to neutron fluences beyond the design operation period

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Shtrombakh, Ya. I.; Nikolaev, Yu. A.; Pechenkin, V. A.

    2005-06-01

    Electron-microscopy and fractographic studies of the surveillance specimens from base and weld metals of VVER-440/213 reactor pressure vessel (RPV) in the original state and after irradiations to different fast neutron fluences from ˜5 × 10 23 n m -2 ( E > 0.5 MeV) up to over design values have been carried out. The maximum specimens irradiation time was 84 480 h. It is shown that there is an evolution in radiation-induced structural behavior with radiation dose increase, which causes a change in relative contribution of the mechanisms responsible for radiation embrittlement of RPV materials. Particularly, radiation coalescence of copper-enriched precipitates and extensive density increase of dislocation loops was observed. Increase in dislocation loop density was shown to provide the dominant contribution to radiation hardening at the late irradiation stages (after reaching double the design end-of-life neutron fluence of ˜4 × 10 24 n m -2). The fracture mechanism of the base metal at those stages was observed to change from transcrystalline to intercrystalline.

  20. Mechanical characteristics and swelling of austenitic Fe-Cr-Mn steels irradiated in the SM-2 and BOR-60 reactors. [0. 4C-12Cr-19Mn-2Ni-Mo-N; 0. 4C-12Cr-14Mn-5Ni-Mo-2Al-B; 0. 4C-17Cr-17Mn-Cu-Mo-Nb-N; Fe-Cr-Ni steel: 0. 8C-16Cr-15Ni-3Mo-Nb; 316; 304

    Energy Technology Data Exchange (ETDEWEB)

    Shamardin, V.K.; Bulanova, T.M.; Neustroev, V.S. (Lenin (V.I.) Research Inst. of Atomic Reactors, Dimitrovgrad (USSR)); Ivanov, L.I.; Djomina, E.V.; Platov, Yu.M. (AN SSSR, Moscow (USSR). A.A. Baikov Inst. of Metallurgy)

    Three types of austenitic Fe-Cr-Mn stainless steels were irradiated simultaneously with Fe-Cr-Ni austenitic steel at temperatures from 400 to 800deg C in the mixed spectrum of the high flux SM-2 reactor to 10 dpa and 700 appm of He and in the BOR-60 reactor to 60 dpa without He generation. The paper presents the swelling and mechanical properties of steels irradiated in the BOR-60 and SM-2 as a function of the concentration of transmuted He and the value of atomic displacement. (orig.).

  1. A facile preparation route for netlike microstructures on a stainless steel using an ethanol-mediated femtosecond laser irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Bian, Hao; Yang, Qing; Liu, Hewei; Chen, Feng, E-mail: chenfeng@mail.xjtu.edu.cn; Du, Guangqing; Si, Jinhai; Hou, Xun

    2013-03-01

    Netlike or porous microstructures are highly desirable in metal implants and biomedical monitoring applications. However, realization of such microstructures remains technically challenging. Here, we report a facile and environmentally friendly method to prepare netlike microstructures on a stainless steel by taking the full advantage of the liquid-mediated femtosecond laser ablation. An unordered netlike structure and a quasi-ordered array of holes can be fabricated on the surface of stainless steel via an ethanol-mediated femtosecond laser line-scan method. SEM analysis of the surface morphology indicates that the porous netlike structure is in the micrometer scale and the diameter of the quasi-ordered holes ranges from 280 nm to 320 nm. Besides, we find that the obtained structures are tunable by altering the laser processing parameters especially scanning speed. - Highlights: Black-Right-Pointing-Pointer A fabrication method of an unordered netlike structure and a quasi-ordered array of holes on metallic surface is developed. Black-Right-Pointing-Pointer The porous netlike structure is in the micrometer scale. Black-Right-Pointing-Pointer The diameter of the quasi-ordered holes ranges from 280 nm to 320 nm. Black-Right-Pointing-Pointer The obtained structures are tunable by altering the laser scanning speed.

  2. Molecular desorption of stainless steel vacuum chambers irradiated with 4.2  MeV/u lead ions

    Directory of Open Access Journals (Sweden)

    E. Mahner

    2003-01-01

    Full Text Available In preparation for the heavy ion program of the Large Hadron Collider at CERN, accumulation and cooling tests with lead ion beams have been performed in the Low Energy Antiproton Ring. These tests have revealed that due to the unexpected large outgassing of the vacuum system, the dynamic pressure of the ring could not be maintained low enough to reach the required beam intensities. To determine the actions necessary to lower the dynamic pressure rise, an experimental program has been initiated for measuring the molecular desorption yields of stainless steel vacuum chambers by the impact of 4.2  MeV/u lead ions with the charge states +27 and +53. The test chambers were exposed either at grazing or at perpendicular incidence. Different surface treatments (glow discharges, nonevaporable getter coating are reported in terms of the molecular desorption yields for H_{2}, CH_{4}, CO, Ar, and CO_{2}. Unexpected large values of molecular yields per incident ion up to 2×10^{4} molecules/ion have been observed. The reduction of the ion-induced desorption yield due to continuous bombardment with lead ions (beam cleaning has been investigated for five different stainless steel vacuum chambers. The implications of these results for the vacuum system of the future Low Energy Ion Ring and possible remedies to reduce the vacuum degradation are discussed.

  3. Impurity content of reduced-activation ferritic steels and a vanadium alloy

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Grossbeck, M.L.; Bloom, E.E. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    Inductively coupled plasma mass spectrometry was used to analyze a reduced-activation ferritic/martensitic steel and a vanadium alloy for low-level impurities that would compromise the reduced-activation characteristics of these materials. The ferritic steel was from the 5-ton IEA heat of modified F82H, and the vanadium alloy was from a 500-kg heat of V-4Cr-4Ti. To compare techniques for analysis of low concentrations of impurities, the vanadium alloy was also examined by glow discharge mass spectrometry. Two other reduced-activation steels and two commercial ferritic steels were also analyzed to determine the difference in the level of the detrimental impurities in the IEA heat and steels for which no extra effort was made to restrict some of the tramp impurities. Silver, cobalt, molybdenum, and niobium proved to be the tramp impurities of most importance. The levels observed in these two materials produced with present technology exceeded the limits for low activation for either shallow land burial or recycling. The chemical analyses provide a benchmark for the improvement in production technology required to achieve reduced activation; they also provide a set of concentrations for calculating decay characteristics for reduced-activation materials. The results indicate the progress that has been made and give an indication of what must still be done before the reduced-activation criteria can be achieved.

  4. Corrosion behavior of EUROFER steel in flowing eutectic Pb-17Li alloy

    Energy Technology Data Exchange (ETDEWEB)

    Konys, J. E-mail: juergen.konys@imf.fzk.de; Krauss, W.; Voss, Z.; Wedemeyer, O

    2004-08-01

    Reduced-activation-ferritic-martensitic (RAFM) steels are considered for application in fusion technology as structural materials. The alloy EUROFER 97 was developed on the basis of the experience gained with steels of the OPTIFER, MANET and F82H-mod. type. These alloys will be in contact with the liquid breeder Pb-17Li and their corrosion behavior is of significance for their successful application. Corrosion tests of EUROFER 97 in flowing Pb-17Li at 480 deg. C were performed up to about 12 000 h to evaluate the kinetics of the dissolution attack. The exposed samples were analysed by metallography and scanning electron microscopy (SEM) with EDX. The results show that EUROFER 97 is attacked by flowing liquid Pb-17Li with a flow velocity of about 0.3 m/s similar to the earlier examined steels and that the typical steel elements are dissolved. The observed attack is of uniform type with values of about 90 {mu}m/year. The corrosion rate is a somewhat smaller for EUROFER compared to the other RAFM steels but with equal activation energy.

  5. Highly antibacterial activity of N-doped TiO{sub 2} thin films coated on stainless steel brackets under visible light irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Shuai; Liu, Bo; Fan, Lingying; Yue, Ziqi [Department of Orthodontics, School of Stomatology, Lanzhou University, Lanzhou 730000 (China); Liu, Bin [Department of Materials Science, School of Physical Science and Technology, Lanzhou University, Lanzhou 730000 (China); Cao, Baocheng, E-mail: caobch@lzu.edu.cn [Department of Orthodontics, School of Stomatology, Lanzhou University, Lanzhou 730000 (China)

    2014-08-01

    In this study, the radio frequency (RF) magnetron sputtering method was used to prepare a TiO{sub 2} thin film on the surface of stainless steel brackets. Eighteen groups of samples were made according to the experimental parameters. The crystal structure and surface morphology were characterized by X-ray diffraction, and scanning electron microscopy, respectively. The photocatalytic properties under visible light irradiation were evaluated by measuring the degradation ratio of methylene blue. The sputtering temperature was set at 300 °C, and the time was set as 180 min, the ratio of Ar to N was 30:1, and annealing temperature was set at 450 °C. The thin films made under these parameters had the highest visible light photocatalytic activity of all the combinations of parameters tested. Antibacterial activities of the selected thin films were also tested against Lactobacillus acidophilus and Candida albicans. The results demonstrated the thin film prepared under the parameters above showed the highest antibacterial activity.

  6. Highly antibacterial activity of N-doped TiO2 thin films coated on stainless steel brackets under visible light irradiation

    Science.gov (United States)

    Cao, Shuai; Liu, Bo; Fan, Lingying; Yue, Ziqi; Liu, Bin; Cao, Baocheng

    2014-08-01

    In this study, the radio frequency (RF) magnetron sputtering method was used to prepare a TiO2 thin film on the surface of stainless steel brackets. Eighteen groups of samples were made according to the experimental parameters. The crystal structure and surface morphology were characterized by X-ray diffraction, and scanning electron microscopy, respectively. The photocatalytic properties under visible light irradiation were evaluated by measuring the degradation ratio of methylene blue. The sputtering temperature was set at 300 °C, and the time was set as 180 min, the ratio of Ar to N was 30:1, and annealing temperature was set at 450 °C. The thin films made under these parameters had the highest visible light photocatalytic activity of all the combinations of parameters tested. Antibacterial activities of the selected thin films were also tested against Lactobacillus acidophilus and Candida albicans. The results demonstrated the thin film prepared under the parameters above showed the highest antibacterial activity.

  7. Molecular desorption of stainless steel vacuum chambers irradiated with 4.2 MeV/u lead ions

    CERN Document Server

    Mahner, E; Laurent, Jean Michel; Madsen, N

    2003-01-01

    In preparation for the heavy ion program of the Large Hadron Collider (LHC) at CERN, accumulation and cooling tests with lead ion beams have been performed in the Low Energy Antiproton Ring (LEAR). These tests have revealed that due to the unexpected large outgassing of the vacuum system, the dynamic pressure of the ring could not be maintained low enough to reach the required beam intensities. To determine the actions necessary to lower the dynamic pressure rise, an experimental program has been initiated for measuring the molecular desorption yields of stainless steel vacuum chambers by the impact of 4.2 MeV/u lead ions with the charge states +27 and +53. The test chambers were exposed either at grazing or at perpendicular incidence. Different surface treatments (glow-discharges, non-evaporable getter coating) are reported in terms of the molecular desorption yields for H2, CH4, CO, Ar and CO2. Unexpected large values of molecular yields per incident ion up to 2 104 molecules/ion have been observed. The red...

  8. Activation and Dose Rate Analysis of 316 Stainless Steel

    Institute of Scientific and Technical Information of China (English)

    XU; Zhi-long; SUN; Zheng; LIU; Xing-min; WAN; Hai-xia

    2012-01-01

    <正>In order to conduct research on 316 stainless steel to be used in reactors, neutron activation during irradiation and dose rate after irradiation in China Experiment Fast Reactor (CEFR) are calculated and analyzed. Based on 1 g of 316 stainless steel specimen, analysis on the activity of 316 stainless steel irradiated

  9. Study on Anti-ablation Property of CrNiMo Steel Irradiated by Laser%CrNiMo钢激光淬火处理的抗烧蚀性能研究

    Institute of Scientific and Technical Information of China (English)

    王文凯; 明学勤

    2009-01-01

    通过激光淬火表面强化技术,在CrNiMo钢表面形成马氏体相变的表面强化薄层.采用小电流累积烧蚀方法分解了材料烧蚀过程;对比分析相同烧蚀条件下CrNiMo钢在激光处理与未经激光处理的表面烧蚀形貌及其烧蚀质量差.结果表明:激光淬火确实能提高材料的抗烧蚀性能.%The thin strengthening layer on the CrNiMo steel surface was prepared by laser surface hardening. The ablation process of the steel was decomposed by the method of accumulated ablation with low current. The ablation surface and mass difference of the original steel and the steel irradiated by laser were analyzed. The results show that the CrNiMo steel through laser guenching indeed promotes the anti-ablation properties of the material.

  10. Void swelling and microstructure evolution at very high damage level in self-ion irradiated ferritic-martensitic steels

    Science.gov (United States)

    Getto, E.; Sun, K.; Monterrosa, A. M.; Jiao, Z.; Hackett, M. J.; Was, G. S.

    2016-11-01

    The void swelling and microstructure evolution of ferritic-martensitic alloys HT9, T91 and T92 were characterized following irradiation with Fe++ ions at 460 °C to damage levels of 75-650 displacements per atom with 10 atom parts per million pre-implanted helium. Steady state swelling rate of 0.033%/ dpa was determined for HT9, the least swelling resistant alloy, and 0.007%/ dpa in T91. In T91, resistance was due to suppression of void nucleation. Swelling resistance was greatest in T92, with a low density (∼1 × 1020 m-3) of small voids that had not grown appreciably, indicating suppression of nucleation and growth. Additional heats of T91 indicated that alloy composition was not the determining factor of swelling resistance. Carbon and chromium-rich M2X precipitates formed at 250 dpa and were correlated with decreased nucleation in T91 and T92, but did not affect void growth in HT9. Dislocation and G-phase microstructure evolution was analyzed up to 650 dpa in HT9.

  11. 核能系统压力容器辐照脆化机制及其影响因素%IRRADIATION EMBRITTLEMENT MECHANISMS AND RELEVANT INFLUENCE FACTORS OF NUCLEAR REACTOR PRESSURE VESSEL STEELS

    Institute of Scientific and Technical Information of China (English)

    李正操; 陈良

    2014-01-01

    Nuclear reactor pressure vessel is the irreplaceable component of the nuclear power plant and its integrity is one of the key issues of any nuclear power plant for long term operations.Various nanofeatures,including solute clusters,matrix damage and grain boundary segregation formed in reactor pressure vessel steels in the face of neutron irradiation.These ultrafine microstructural features lead to an increase in the ductile brittle transition temperature as is the measure used to describe the irradiation embrittlement.The balance of features depends on the composition of the reactor pressure vessel steels and the irradiation conditions.This paper reviews the current phenomenological knowledge and understanding of the basic mechanisms and relevant influence factors for irradiation embrittlement of nuclear reactor pressure vessel steels.To be specific,the formation and evolution processes of the embrittling features are presented.Also,the influences of material variables,such as copper,nickel and manganese contents on irradiation embrittlement and those of irradiation variables,such as neutron flux and post irradiation annealing are summarized.In addition,fundamental research issues that remain to be addressed are briefly pointed out.%核反应堆压力容器作为核电站不可更换的关键性设备,其设备完整性对核电站的安全运行起着至关重要的作用.在辐照条件下,反应堆压力容器钢中会形成一系列微结构缺陷,包括溶质沉淀、基体损伤和脆性元素的晶界偏聚等,导致材料的韧脆性转变温度升高,产生辐照脆化效应.而压力容器钢的成分和辐照条件决定了各种微结构对辐照脆化的贡献大小.本文主要针对核能系统压力容器辐照脆化机制及其影响因素进行了综述,总结讨论了这些微结构的形成机制及溶质元素、辐照通量和辐照后退火对这些微结构和材料机械性能的影响,并指出了存在的问题和未来的研究方向.

  12. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects; Mecanismes de fragilisation sous irradiation aux neutrons d'alliages modeles ferritiques et d'un acier de cuve: amas de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

    The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)

  13. Results from the irradiation of stainless steel and copper by 23 MeV γ-quanta in the atmosphere of molecular deuterium at a pressure of 2 kbar

    Science.gov (United States)

    Didyk, A. Yu.; Wisniewski, R.

    2014-05-01

    Metal samples were arranged inside a "finger-type" high-pressure chamber (DHPC-FT) filled by deuterium. They were two aluminum rods, a copper rod, two specimens of homogeneous YMn2 alloy, and a stainless steel wire. The pressure of molecular deuterium in DHPC-FT was about 2 kbar. The samples were irradiated by braking γ-quanta at a threshold energy of 23 MeV. All the samples were studied using scanning electron microscopy (SEM) and X-ray (roentgen) microelement probe analysis (RMPA). Considerable changes in the surface structure and elemental composition were found for the samples of copper, aluminum, YMn2 alloy, and stainless steel. Unusual results were analyzed in detail and compared with the earlier data.

  14. Compatibility of reduced activation ferritic/martensitic steel specimens with liquid Na and NaK in irradiation rig of IFMIF

    OpenAIRE

    2005-01-01

    In the high flux region of the International Fusion Materials Irradiation Facility (IFMIF), the neutron irradiation damage for iron-based alloys will exceed 20 dpa/ year. An accurate specimen temperature measurement under a large amount of nuclear heating is a key issue but the change of heat transfer of gap between irradiation specimens and specimen holder during irradiation test is inevitable, if gap is filled with an inert gas and temperature is monitored by a thermocouple buried in the sp...

  15. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  16. 3D finite element and experimental study of the size requirements for measuring toughness on tempered martensitic steels

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, P. [Fusion Technology-Materials, CRPP-EPFL, Association EURATOM-Confederation Suisse, 5232 Villigen PSI (Switzerland)], E-mail: pablo.mueller@psi.ch; Spaetig, P. [Fusion Technology-Materials, CRPP-EPFL, Association EURATOM-Confederation Suisse, 5232 Villigen PSI (Switzerland)

    2009-06-01

    The fracture properties of the tempered martensitic steel Eurofer97, which is among the main candidates for fusion power plant structural applications, were studied with two sizes of pre-cracked compact specimens (0.35T C(T) and 0.87T C(T)). The fracture toughness behavior was characterized within the temperature range -80 to -40 deg. C. The ductile-to-brittle transition reference temperature, as defined in the ASTM standard E1921, was around T{sub 0} {approx} -75 deg. C. At -60 deg. C, it was found that two sets of toughness data obtained with 0.35T and 0.87T C(T) specimens are not consistent with the size adjustments recommended in the ASTM standard. It was then shown that the underlying reason of this inconsistency is an inappropriate specimen size limit of the ASTM standard for this type of steel. From published fracture toughness data on the tempered martensitic steel F82H steel, similar results were also highlighted. 3D finite elements simulations of the compact specimens were performed to compare the stresses and deformations at the onset of fracture. A local approach model based on the attainment of a critical stress and a critical volume was used to study the constraint loss phenomenon. Within the framework of this model, the strong toughness increase by reducing the specimen size could be satisfactorily explained.

  17. 3D finite element and experimental study of the size requirements for measuring toughness on tempered martensitic steels

    Science.gov (United States)

    Mueller, P.; Spätig, P.

    2009-06-01

    The fracture properties of the tempered martensitic steel Eurofer97, which is among the main candidates for fusion power plant structural applications, were studied with two sizes of pre-cracked compact specimens (0.35T C(T) and 0.87T C(T)). The fracture toughness behavior was characterized within the temperature range -80 to -40 °C. The ductile-to-brittle transition reference temperature, as defined in the ASTM standard E1921, was around T0 ≈ -75 °C. At -60 °C, it was found that two sets of toughness data obtained with 0.35T and 0.87T C(T) specimens are not consistent with the size adjustments recommended in the ASTM standard. It was then shown that the underlying reason of this inconsistency is an inappropriate specimen size limit of the ASTM standard for this type of steel. From published fracture toughness data on the tempered martensitic steel F82H steel, similar results were also highlighted. 3D finite elements simulations of the compact specimens were performed to compare the stresses and deformations at the onset of fracture. A local approach model based on the attainment of a critical stress and a critical volume was used to study the constraint loss phenomenon. Within the framework of this model, the strong toughness increase by reducing the specimen size could be satisfactorily explained.

  18. 高温氘离子辐照对低活化钢微观结构的影响%Microstructure changes of reduced activation steel after deuterium ion irradiation at high temperature

    Institute of Scientific and Technical Information of China (English)

    刘平平; 詹倩; 赵明忠; 万发荣; 大貫惣明

    2014-01-01

    材料问题是可控核聚变能否实现商业应用从而解决人类能源问题的“瓶颈”之一。低活化铁素体/马氏体(RAFM)钢具有良好的抗辐照性能,被普遍认为是新一代聚变反应堆的候选结构材料之一。在聚变堆环境下,材料不仅会受到高能中子辐照而且氘氚也可能进入材料中。为了研究氘离子以及辐照对低活化钢的微观结构的影响,采用 CLAM钢(一种 RAFM钢)和FeCr模型合金,在500℃下进行58 keV氘离子辐照,利用高分辨透射电镜对比分析辐照前后材料微观结构的变化,研究辐照及氘离子对低活化钢的影响。结果表明:高温氘离子辐照不仅在材料中产生大量的缺陷和缺陷集团,同时还可能产生辐照诱导析出。而CLAM钢中原有的析出物经高温离子辐照后并没有发生非晶化,对其原因进行了讨论。%A good understanding on the effect of deuterium which exists in the environment of fusion reactions and irradiation is important to develop structural materials in fusion reactors and for a safe design and operation of innovative nuclear systems.Reduced activation ferritic/matensitic (RAFM)steel is considered as the candi-date of structural materials for fusion power reactors due to their excellent resistance to swelling under irradia-tion and low cost.To investigate the effects of deuterium ion and irradiation on the microstructure of RAFM steel,CLAM steel,one kind of RAFM steel,and Fe-10Cr model alloy have been irradiated by 58 keV deuteri-um ion at 500 ℃.The microstructural changes were investigated by high-resolution transmission electron micro-scope (HRTEM).The results show that lots of defects such as dislocation loops have been induced.Precipitates have also been induced after deuterium ion irradiation at high temperature.According to the HRTEM image of M23 C6 in irradiated CLAM steel,no obvious amorphization has been observed.This complex phenomenon was discussed.

  19. An austenitic steel for fuel cladding tubes and core components of LMFBR`s with high ductility after neutron irradiation; Ein austenitischer Stahl fuer Huellrohre und Kernkomponenten natriumgekuehlter Brueter mit hoher Duktilitaet nach Neutronenbestrahlung

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, L.; Kempe, H.

    1994-06-01

    Two heats of an austenitic stainless steel with different priority concerning the resistance against Helium-embrittlement (B801) and void-swelling (F218) had been developed and tested as a material for fuel rod claddings and core components of liquid metal fast breeder reactors. The two steels show a ductility five times higher than the reference steel 1.4970 in tensile - and creep-rupture-tests after irradiation in reactors with fast and mixed neutron flux respectively. Just so the swelling resistance had been confirmed up to 40 dpa. Checked claddings of the heat F218 in the dimensions 6x0.38 mm, 6.55x0.45 mm and 7.6x0.5 mm are available for pin- and bundle irradiation experiments. (orig.) [Deutsch] Im Rahmen der Entwicklung austenitischer Staehle als Werkstoffe fuer Huellrohre und Kernkomponenten Schneller Natriumgekuehlter Brutreaktoren wurden zwei Chargen mit unterschiedlicher Prioritaet fuer ihre Widerstandsfaehigkeit gegen Heliumversproedung (B801) und Porenschwellen (F218) konzipiert und untersucht. Beide Staehle zeigten nach Bestrahlung in Reaktoren mit schnellem bzw. gemischtem Neutronenfluss sowohl im Warmzugversuch als auch im Zeitstandversuch eine Duktilitaet, die um den Faktor 5 hoeher liegt als die des Referenzstahles 1.4970. Fuer beide Staehle konnte die Schwellresistenz bis 40 dpa Neutronenbestrahlung nachgewiesen werden. Fuer Brennstab- und Buendelbestrahlungsexperimente stehen gepruefte Huellrohre der Charge F218 mit den Abmessungen 6x0.38 mm, 6.55x0.45 mm und 7.6x0.5 mm zur Verfuegung. (orig.)

  20. Metallurgical properties study of reduced activation ferritic/martensitic steels for their application in a fusion reactor; Estudio de las propiedades metalurgicas de los aceros martensticos de activacion reducida para su aplicacion en los reactores de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez Paredes, M. P.

    2007-07-01

    In this work, the metallurgical characterization of two reduced activation ferritic/martensitic steel of last generation, called F-82H-mod and Eurofer'97 has been performed. the aim of this study is to contribute to the knowledge and understanding of the degradation mechanisms and ageing that these materials can suffer during operation at high temperatures with or without the application of the load (creep). In order to fulfill these objectives, the investigations have been carried out in both materials on as-received condition and after thermal ageing treatments in the temperature range from 300 degree centigree to 600 degree centigree for periods up to 1000 h. To achieve the objectives cited, microstructural (optical, SEM, phase extraction, X-ray diffraction and TEM) and mechanical (hardness, tensile, Charpy, Fracture toughness and creep) investigations have been performed. (Author) 58 refs.

  1. Experimental investigation of high He/dpa microstructural effects in neutron irradiated B-alloyed Eurofer97 steel by means of small angle neutron scattering (SANS and electron microscopy

    Directory of Open Access Journals (Sweden)

    R. Coppola

    2016-12-01

    Full Text Available High He/dpa microstructural effects have been investigated, by means of small-angle neutron scattering (SANS and transmission electron microscopy (TEM, in B-alloyed ferritic/martensitic steel Eurofer97-1 (0.12 C, 9 Cr, 0.2V, 1.08W wt%, B contents variable between 10 and 1000ppm, neutron irradiated at the High Flux Reactor of the JRC-Petten at temperatures between 250 °C and 450 °C, up do a dose level of 16 dpa. Under these irradiation parameters, B activation is expected to produce corresponding helium contents variable between 80 and 5600appm, with helium bubble distributions relevant for the technological applications. The SANS measurements were carried out under magnetic field to separate nuclear and magnetic SANS components; a reference, un-irradiated sample was also measured to evaluate as accurately as possible the genuine effect of the irradiation on the microstructure. Increasing the estimated helium content from 400 to 5600appm, the analysis of the SANS cross-sections yields an increase in the volume fraction, attributed to helium bubbles, of almost one order of magnitude (from 0.007 to 0.038; furthermore, the difference between nuclear and magnetic SANS components is strongly reduced. These results are discussed in correlation with TEM observations of the same samples and are tentatively attributed to the effect of drastic microstructural changes in Eurofer97-1 for high He/dpa ratio values, possibly relating to the dissolution of large B-carbides due to transmutation reactions.

  2. Fe-15Ni-13Cr austenitic stainless steels for fission and fusion reactor applications - Part III: Phase stability during heavy ion irradiation

    Science.gov (United States)

    Lee, E. H.; Mansur, L. K.

    2000-01-01

    The phase stability in Fe-15Ni-13Cr alloys was investigated as a function of minor alloying additions after 4 MeV Ni ion irradiation at 948 K. The results showed that the stability of precipitate phases was dictated mainly by the defects produced by radiation damage and preferential segregation of Si and Ni at defects. In addition, radiation enhanced diffusion and cascade induced dissolution and mixing allowed kinetically sluggish phases to form rapidly under irradiation. These radiation effects caused an enhancement, retardation, or modification of thermal phases, and formation of new phases. The relative stability of precipitate phases varied sensitively with alloy composition. The roles of each alloying element on phase stability and the impact of radiation on the mechanisms of phase evolution were systematically studied and documented. The knowledge obtained from this work provides guidelines for designing alloys that lead to develop desired precipitate microstructures under irradiation.

  3. High energy X-ray diffraction study of the relationship between the macroscopic mechanical properties and microstructure of irradiated HT-9 steel

    Science.gov (United States)

    Tomchik, C.; Almer, J.; Anderoglu, O.; Balogh, L.; Brown, D. W.; Clausen, B.; Maloy, S. A.; Sisneros, T. A.; Stubbins, J. F.

    2016-07-01

    Samples harvested from an HT-9 fuel test assembly (ACO-3) irradiated for six years in the Fast Flux Test Facility (FFTF) reaching 2-147 dpa at 382-504 °C were deformed in-situ while collecting high-energy X-ray diffraction data to monitor microstructure evolution. With the initiation of plastic deformation, all samples exhibited a clear load transfer from the ferrite matrix to carbide particulate. This behavior was confirmed by modeling of the control material. The evolution of dislocation density in the material as a result of deformation was characterized through full pattern line profile analysis. The dislocation densities increased substantially after deformation, the level of dislocation evolution observed was highly dependent upon the irradiation temperature of the sample. Differences in both the yield and hardening behavior between samples irradiated at higher and lower temperatures suggest the existence of a transition in tensile behavior at an irradiation temperature near 420 °C dividing regions of distinct damage effects.

  4. A study of the neutron irradiation effects on the susceptibility to embrittlement of A316L and T91 steels in lead-bismuth eutectic

    Energy Technology Data Exchange (ETDEWEB)

    Sapundjiev, D. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium)]. E-mail: danislav.sapundjiev@sckcen.be; Al Mazouzi, A. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium); Van Dyck, S. [TCH, SCK-CEN, Boeretang 200, Mol, B-2400 (Belgium)

    2006-09-15

    The effects of neutron irradiation on the susceptibility to liquid metal embrittlement of two primary selected materials for MYRRHA project an accelerator driven system (ADS), was investigated by means of slow strain rate tests (SSRT). The latter were carried out at 200 deg. C in nitrogen and in liquid Pb-Bi at a strain rate of 5 x 10{sup -6} s{sup -1}. The small tensile specimens were irradiated at the BR-2 reactor in the MISTRAL irradiation rig at 200 deg. C for 3 reactor cycles to reach a dose of about 1.50 dpa. The SSR tests were carried out under poor and under dissolved oxygen conditions ({approx}1.5 x 10{sup -12} wt% dissolved oxygen) which at this temperature will favour formation of iron and chromium oxides. Although both materials differ in structure (fcc for A316L against bcc for T91), their flow behaviour in contact with liquid lead bismuth eutectic before and after irradiation is very similar. Under these testing conditions none of them was found susceptible to liquid metal embrittlement (LME)

  5. Estimates of time-dependence fatigue behavior of type 316 stainless steel subject to irradiation damage in fast breeder and fusion power reactor systems. [Fluence 1--2. 63 x 10/sup 26/ n/m/sup 2/ (E > 0. 1 MeV/ at 593/sup 0/C

    Energy Technology Data Exchange (ETDEWEB)

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

    Cyclic lives obtained from strain controlled fatigue tests at 593/sup 0/C from specimens irradiated to a fluence of 1 to 2.63 x 10/sup 26/ n/m/sup 2/ (E greater than 0.1 MeV) were compared to predictions based on the method of strainrange partitioning. 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. 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 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/sup 2/ were used. Results, although conjectural because of the many assumptions, tended to show that 20 percent cold-worked type 316 stainless steel could be used as a first wall material meeting a 7.5 to 8.5 MW-year/m/sup 2/ lifetime goal provided the neutron wall loading does not exceed more than about 2 MW/m/sup 2/. Results were obtained for an air environment, and it is expected that the actual vacuum environment will extend lifetime beyond 10 MW-year/m/sup 2/.

  6. Ion-stimulated gas desorption yields of coated (Au, Ag, Pd) stainless steel vacuum chambers irradiated with 4.2 MeV/u lead ions

    CERN Document Server

    Mahner, E; Küchler, D; Malabaila, M; Taborelli, M

    2005-01-01

    The ion-induced desorption experiment, installed in the CERN Heavy Ion Accelerator (LINAC 3), has been used to measure molecular desorption yields for 4.2 MeV/u lead ions impacting on different accelerator-type vacuum chambers. In order to study the effect of the surface oxide layer on the gas desorption, gold-, silver-, and palladium-coated 316LN stainless steel chambers and similarly prepared samples were tested for desorption at LINAC 3 and analysed for chemical composition by X-ray Photoemission Spectroscopy (XPS). The large effective desorption yield of 2 x 10**4 molecules/ion, previously measured for uncoated, vacuum fired stainless steel, was reduced after noble metal coating by up to 2 orders of magnitude. In addition, the effectiveness of beam scrubbing with heavy ions and the consequence of a subsequent venting on the desorption yields of a beam-scrubbed vacuum chamber are described. Practical consequences for the vacuum system of the future Low Energy Ion Ring (LEIR) are discussed.

  7. Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki [Japan Atomic Energy Agency (JP)] (and others)

    2007-07-01

    Reduced activation ferritic/martensitic steels (RAFMs) are recognized as the primary candidate structural materials for fusion blanket systems, as it has they have been developed based on massive industrial experience of ferritic/martensitic steel replacing Mo and Nb of high chromium heat resistant martensitic steels (such as modified 9Cr-1Mo) with W and Ta, respectively. F82H and JLF-1 are RAFMs, which have been developed and studied in Japan and the various effects of irradiation were reported. F82H is designed with emphasis on high temperature property and weldability, and was provided and evaluated in various countries as a part of the IEA fusion materials development collaboration. The JAEA/US collaboration program also has been conducted with the emphasis on irradiation effects of F82H. Now, among the existing database for RAFMs the most extensive one is that for F82H. The objective of this paper is to review the R and D status of F82H and to identify the key technical issues for the fabrication of ITER Test Blanket Module (TBM) suggested from the recent achievements in Japan. It is desirable to make the status of RAFMs equivalent to commercial steels to use RAFMs as the ITER-TBM structural material. This would require demonstrating the reproducibility and weldability as well as providing the database. The excellent reproducibility of F82H has been demonstrated with four 5-ton-heats, and two of them were provided as F82H-IEA heats. It has been also proved that F82H could be provided as plates (thickness of 1.5 to 55 mm), pipes and rectangular tubes. It is also important to have the excellent weldability as the TBM has about 300m length of weld line, and it was proved through TIG, EB and YAG weld test performed in air atmosphere. Various mechanical and microstructural data have been accumulated including long-term tests such as creep rupture tests and aging tests. Although F82H is a well-perceived RAFM as the ITER-TBM structural material, some issues are

  8. Steel making

    CERN Document Server

    Chakrabarti, A K

    2014-01-01

    "Steel Making" is designed to give students a strong grounding in the theory and state-of-the-art practice of production of steels. This book is primarily focused to meet the needs of undergraduate metallurgical students and candidates for associate membership examinations of professional bodies (AMIIM, AMIE). Besides, for all engineering professionals working in steel plants who need to understand the basic principles of steel making, the text provides a sound introduction to the subject.Beginning with a brief introduction to the historical perspective and current status of steel making together with the reasons for obsolescence of Bessemer converter and open hearth processes, the book moves on to: elaborate the physiochemical principles involved in steel making; explain the operational principles and practices of the modern processes of primary steel making (LD converter, Q-BOP process, and electric furnace process); provide a summary of the developments in secondary refining of steels; discuss principles a...

  9. Steel Spring

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Tarnished Hebei Iron and Steel Group regains chance to shine A lthough it is too early to tell whether the steel-making sector has emerged [from its gloom, a big divide is openling between China’s large and small producers. While most of the marginal players are still reeling from a market contagion, steel titans like the Shanghai-based Baosteel

  10. Results of investigations regarding the physical and mechanical properties of the martensitic 9% Cr steel EUROFER '97; Ergebnisse von Charakterisierungsuntersuchungen zu physikalischen und mechanischen Eigenschaften des martensitischen 9% Cr-Stahles EUROFER '97

    Energy Technology Data Exchange (ETDEWEB)

    Schirra, M.; Falkenstein, A.; Graf, P.; Heger, S.; Kempe, H.; Lindau, R.; Zimmermann, H.

    2002-04-01

    Following the history of the development activities from conventional martensitic 12% Cr steel, MANET and OPTIFER up to low-activated EUROFER (RAFM steel), the results obtained from experiments on rods of 100 mm diameter and plates of 14 mm shall be presented for a basic characterization. The physical and mechanical properties shall be compared with those of OPTIFER-1W and the F82H-mod 2% W steel. To determine the conversion behavior, a continuous cct diagram was plotted. Hardening experiments in the temperature range of 850 - 1120 C illustrated the range of maximum hardness as well as grain size development. The notch impact behavior was described for various heat treatments and sample types at test temperatures ranging from +60 to -100 C. Tensile strengths were determined for various heat treatments at temperatures ranging from room temperature to 700 C. Aging due to a long-term heat treatment was investigated by means of stabilization annealing experiments. Creep rupture strength and creeping were investigated in the temperature range of 450 - 650 C. So far, a duration of up to about 15 000 h has been covered by the experiments. (orig.)

  11. On the crystal structure of Cr2N precipitates in high-nitrogen austenitic stainless steel. II. Order-disorder transition of Cr2N during electron irradiation.

    Science.gov (United States)

    Lee, Tae Ho; Kim, Sung Joon; Takaki, Setsuo

    2006-04-01

    The crystal structure and order-disorder transition of Cr2N were investigated utilizing transmission electron microscopy (TEM). Based on the analyses of selected-area diffraction (SAD) patterns, the crystal structure of the ordered Cr2N superstructure was confirmed to be trigonal (P31m), characterized by three sets of superlattice reflections (001), ((11/33)0) and ((11/33)1). During electron irradiation, the superlattice reflections gradually disappeared in the regular sequence (001), ((11/33)0) and ((11/33)1), indicating that the order-disorder phase transition of Cr2N occurred. The convergent-beam electron diffraction (CBED) observation revealed that the space group of disordered Cr2N is P6(3)/mmc, which corresponds to an h.c.p. (hexagonal close packed) sublattice of metal atoms with a random distribution of N atoms in six octahedral interstices. The redistribution model of N atoms through the order-disorder transition is discussed based on the characteristics and disappearing sequence of superlattice reflections.

  12. (Irradiation embrittlement of reactor pressure vessels)

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, W.R.

    1990-09-24

    The traveler served as a member of the two-man US Nuclear Regulatory Commission sponsored team who visited the Prometey Complex in Leningrad to assess the potential for expanded cooperative research concerning integrity of the primary pressure boundary in commercial light-water reactors. The emphasis was on irradiation embrittlement, structural analysis, and fracture mechanics research for reactor pressure vessels. At the irradiation seminar in Cologne, presentations were made by German, French, Finnish, Russian, and US delegations concerning many aspects of irradiation of pressure vessel steels. The traveler made presentations on mechanisms of irradiation embrittlement and on important aspects of the Heavy-Section Steel Irradiation Program results of irradiated fracture mechanics tests.

  13. Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

    1999-08-01

    Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

  14. Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

    1999-08-01

    Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

  15. Tool steels

    DEFF Research Database (Denmark)

    Højerslev, C.

    2001-01-01

    On designing a tool steel, its composition and heat treatment parameters are chosen to provide a hardened and tempered martensitic matrix in which carbides are evenly distributed. In this condition the matrix has an optimum combination of hardness andtoughness, the primary carbides provide...... resistance against abrasive wear and secondary carbides (if any) increase the resistance against plastic deformation. Tool steels are alloyed with carbide forming elements (Typically: vanadium, tungsten, molybdenumand chromium) furthermore some steel types contains cobalt. Addition of alloying elements...

  16. Surface cracking and melting of different tungsten grades under transient heat and particle loads in a magnetized coaxial plasma gun

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Y., E-mail: ykikuchi@eng.u-hyogo.ac.jp [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Sakuma, I.; Iwamoto, D.; Kitagawa, Y.; Fukumoto, N.; Nagata, M. [Graduate School of Engineering, University of Hyogo, 2167 Shosha, Himeji, Hyogo 671-2280 (Japan); Ueda, Y. [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

    2013-07-15

    Surface damage of pure tungsten (W), W alloys with 2 wt.% tantalum (W–Ta) and vacuum plasma spray (VPS) W coating on a reduced activation material of ferritic steel (F82H) due to repetitive ELM-like pulsed (∼0.3 ms) deuterium plasma irradiation has been investigated by using a magnetized coaxial plasma gun. Surface cracks appeared on a pure W sample exposed to 10 plasma pulses of ∼0.3 MJ m{sup −2}, while a W–Ta sample did not show surface cracks with similar pulsed plasma irradiation. The energy density threshold for surface cracking was significantly increased by the existence of the alloying element of tantalum. No surface morphology change of a VPS W coated F82H sample was observed under 10 plasma pulses of ∼0.3 MJ m{sup −2}, although surface melting and cracks in the resolidification layer occurred at higher energy density of ∼0.9 MJ m{sup −2}. There was no indication of exfoliation of the W coating from the substrate of F82H after the pulsed plasma exposures.

  17. Surface cracking and melting of different tungsten grades under transient heat and particle loads in a magnetized coaxial plasma gun

    Science.gov (United States)

    Kikuchi, Y.; Sakuma, I.; Iwamoto, D.; Kitagawa, Y.; Fukumoto, N.; Nagata, M.; Ueda, Y.

    2013-07-01

    Surface damage of pure tungsten (W), W alloys with 2 wt.% tantalum (W-Ta) and vacuum plasma spray (VPS) W coating on a reduced activation material of ferritic steel (F82H) due to repetitive ELM-like pulsed (˜0.3 ms) deuterium plasma irradiation has been investigated by using a magnetized coaxial plasma gun. Surface cracks appeared on a pure W sample exposed to 10 plasma pulses of ˜0.3 MJ m-2, while a W-Ta sample did not show surface cracks with similar pulsed plasma irradiation. The energy density threshold for surface cracking was significantly increased by the existence of the alloying element of tantalum. No surface morphology change of a VPS W coated F82H sample was observed under 10 plasma pulses of ˜0.3 MJ m-2, although surface melting and cracks in the resolidification layer occurred at higher energy density of ˜0.9 MJ m-2. There was no indication of exfoliation of the W coating from the substrate of F82H after the pulsed plasma exposures.

  18. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    Science.gov (United States)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-06-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels.

  19. Steel Planning

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    China releases a new plan for the iron and steel industry centered on industrial upgrades The new 12th Five-Year Plan (2011-15) for China’s iron and steel industry, recently released on the website of the Ministry of Industry and Information

  20. APT characterization of high nickel RPV steels

    Science.gov (United States)

    Miller, M. K.; Sokolov, M. A.; Nanstad, R. K.; Russell, K. F.

    2006-06-01

    The microstructures of three high nickel content pressure vessel steels have been characterized by atom probe tomography to investigate the influence of high nickel levels on the response to neutron irradiation of high and low copper pressure vessel steels. The high-nickel, low-manganese, low-copper VVER-1000 weld and forging exhibited lower than predicted levels of embrittlement during neutron irradiation. The Palisades weld exhibits a Δ T41 J of 102 °C which was significantly lower than the value of 154 °C predicted by Reg. Guide 1.99 Rev. 2. Atom probe tomography revealed nickel-, manganese-, and silicon-enriched precipitates in both the VVER-1000 base and weld materials after neutron irradiation. A high number density of copper-, nickel-, manganese-, silicon- and phosphorus-enriched precipitates were observed in the Palisades weld after neutron irradiation. Atom probe tomography also revealed high levels of phosphorus segregation to the dislocations in all three materials.

  1. Irradiation creep of dispersion strengthened copper alloy

    Energy Technology Data Exchange (ETDEWEB)

    Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

    1997-04-01

    Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

  2. Irradiation damage

    Energy Technology Data Exchange (ETDEWEB)

    Howe, L.M

    2000-07-01

    There is considerable interest in irradiation effects in intermetallic compounds from both the applied and fundamental aspects. Initially, this interest was associated mainly with nuclear reactor programs but it now extends to the fields of ion-beam modification of metals, behaviour of amorphous materials, ion-beam processing of electronic materials, and ion-beam simulations of various kinds. The field of irradiation damage in intermetallic compounds is rapidly expanding, and no attempt will be made in this chapter to cover all of the various aspects. Instead, attention will be focused on some specific areas and, hopefully, through these, some insight will be given into the physical processes involved, the present state of our knowledge, and the challenge of obtaining more comprehensive understanding in the future. The specific areas that will be covered are: point defects in intermetallic compounds; irradiation-enhanced ordering and irradiation-induced disordering of ordered alloys; irradiation-induced amorphization.

  3. Food irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Webb, T.

    1986-01-01

    The proposed use of gamma radiation from cobalt 60 and cesium 137 for food irradiation in the United Kingdom is discussed, with particular reference to the possible dangers and disadvantages to the safety and wholesomeness of the food.

  4. [Food irradiation].

    Science.gov (United States)

    Migdał, W

    1995-01-01

    A worldwide standard on food irradiation was adopted in 1983 by Codex Alimentarius Commission of the Joint Food Standard Programme of the Food and Agriculture Organization (FAO) of the United Nations and the World Health Organization (WHO). As a result, 41 countries have approved the use of irradiation for treating one or more food items and the number is increasing. Generally, irradiation is used to: food loses, food spoilage, disinfestation, safety and hygiene. The number of countries which use irradiation for processing food for commercial purposes has been increasing steadily from 19 in 1987 to 33 today. In the frames of the national programme on the application of irradiation for food preservation and hygienization an experimental plant for electron beam processing has been established in Institute of Nuclear Chemistry and Technology. The plant is equipped with a small research accelerator Pilot (19MeV, 1 kW) and an industrial unit Elektronika (10MeV, 10 kW). On the basis of the research there were performed at different scientific institutions in Poland, health authorities have issued permission for irradiation for: spices, garlic, onions, mushrooms, potatoes, dry mushrooms and vegetables.

  5. High Nitrogen Stainless Steel

    Science.gov (United States)

    2011-07-19

    Kiev, 1993. 7. High Nitrogen Steels, edited by M. Kikuchi and Y. Mishima , Vol. 36, No. 7, Iron and Steel Institute of Japan Inernational, Tokyo...the Corrosion of Iron and Steels,” High Nitrogen Steels, edited by M. Kikuchi and Y. Mishima , Vol. 36, No. 7, Iron and Steel Institute of Japan

  6. SANS response of VVER440-type weld material after neutron irradiation, post-irradiation annealing and reirradiation

    OpenAIRE

    Ulbricht, Andreas; Bergner, Frank; Boehmert, Juergen; Valo, Matti; Mathon, Marie-Helene; Heinemann, Andre

    2007-01-01

    Abstract It is well accepted that the reirradiation behaviour of reactor pressure vessel (RPV) steel after annealing can be different from the original irradiation behaviour. We present the first small-angle neutron scattering (SANS) study of neutron irradiated, annealed and reirradiated VVER440-type RPV weld material. The SANS results are analysed both in terms of the size distribution of irradiation-induced defect/solute atom clusters and in terms of the ratio of total and nuclea...

  7. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Ok; Hwang, Mi Jin; Kim, Sung Woo; Hwang, Seong Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples.

  8. Ion-stimulated Gas Desorption Yields of Electropolished, Chemically Etched, and Coated (Au, Ag, Pd, TiZrV) Stainless Steel Vacuum Chambers and St707 Getter Strips Irradiated with 4.2 MeV/u lead ions

    CERN Document Server

    Mahner, E; Küchler, D; Malabaila, M; Taborelli, M

    2005-01-01

    The ion-induced desorption experiment, installed in the CERN Heavy Ion Accelerator LINAC 3, has been used to measure molecular desorption yields for 4.2 MeV/u lead ions impacting under grazing incidence on different accelerator-type vacuum chambers. Desorption yields for H2, CH4, CO, and CO2, which are of fundamental interest for future accelerator applications, are reported for different stainless steel surface treatments. In order to study the effect of the surface oxide layer on the gas desorption, gold-, silver-, palladium-, and getter-coated 316 LN stainless steel chambers and similarly prepared samples were tested for desorption at LINAC 3 and analysed for chemical composition by X-ray Photoemission Spectroscopy (XPS). The large effective desorption yield of 2 x 104 molecules/Pb53+ ion, previously measured for uncoated, vacuum fired stainless steel, was reduced after noble-metal coating by up to 2 orders of magnitude. In addition, pressure rise measurements, the effectiveness of beam scrubbing with le...

  9. Evaluation of impacts of stress triaxiality on plastic deformability of RAFM steel using various types of tensile specimen

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Taichiro, E-mail: kato.taichiro@jaea.go.jp [Japan Atomic Energy Agency, 2-166, Obuchi-omotedate, Rokkasho, Aomori 039-3212 (Japan); Ohata, Mitsuru [Osaka University, 2-1, Yamada-Oka, Suita, Osaka 565-0871 (Japan); Nogami, Shuhei [Tohoku University, 6-6-01-2, Aramaki-aza-Aoba, Aoba-ku, Sendai, Miyagi 980-8579 (Japan); Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166, Obuchi-omotedate, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • The fracture ductility is lower as the stress triaxiality is higher. • Voids of the interrupted RB1 specimen were observed along grain boundaries and expanded parallel to the tensile axis. • Voids of interrupted R0.2 specimen were rounded shape than those of RB1. • The fracture surface of specimens were observed the elongated and the equiaxed dimples. • The decrease of plastic deformability of the notched specimen was caused by the process of voids formation and crack growth due to the effect of plastic constraint of the notch. - Abstract: A case study on a fusion blanket design such as DEMO indicated that there could be some sections with high stress triaxiality, a parameter to evaluate the magnitude of plastic constraint, in the case of plasma disruption or coolant loss accident. Therefore, it is necessary to accurately understand the ductility loss limit of structural material in order to conduct the structural design assessment of the irradiated and embrittled fusion reactor blanket. Tensile tests were conducted by using three kinds of tensile specimen shapes to investigate of the plastic deformability of F82H. From the results, the fracture ductility is lower as the stress triaxiality is higher. Voids of the interrupted RB1 specimen were observed along grain boundaries and expanded parallel to the tensile axis. That of interrupted R0.2 specimen was rounded shape compared with those of RB1. The fracture surface of RB1 and R0.2 specimens were observed the elongated dimples and the equiaxed dimples without so much elongation, respectively. It is considered that the decrease of plastic deformability for the notched specimen was caused by the process of voids formation and crack growth due to the effect of plastic constraint of the notch.

  10. Analysis of Effects in a Low-activation Ferritic/Martensitic Steel by High-energy Heavy-ion Irradiation%一种低活化铁素体/马氏体钢的高能重离子辐照效应研究

    Institute of Scientific and Technical Information of China (English)

    张崇宏; 杨义涛; 宋银; J.S.Jang; 孙友梅; 金运范; 李炳生

    2009-01-01

    Low-activation Ferritic/Martensitic steels are a kind of important structural materials candidate to the application in advanced nuclear energy systems. Possible degradation of properties and even failure in the condition of high-temperature and high helium production due to energetic neutron irradiation in a fusion reactor is a major concern with the application of this kind of materials. In the present work micro-structural evolution in a 9Cr Ferritic/Martensitic steel (T92B) irradiated with 122 MeV ~(20)Ne ions at temperatures between 0.3-0.5 T_m (T_m is the melting point of the material) was investigated with transmission electron microscopy. High concentration voids were observed in the specimens irradiated at high temperatures when the displacement damage dose and Ne concentration exceed a certain level. Preferential formation of voids at lath-boundaries and other grain-boundaries was found. The data of void swellings in 9Cr ferritic/martensitic steels irradiated in different conditions (such as with He-ions, Ne-ions, Fe/He dual beams, fast neutrons, Ni-ions etc. ) were compiled and analyzed based on a classic model of helium bubble formation, and bubble to void transition.%低活化的铁素体/马氏体钢是先进核能装置(如聚变堆)的重要候选结构材料.在聚变堆实际工作环境下,由于高温和高氦产生率引起的材料失效是这类材料面临的一个重要问题.本项研究以兰州重离子加速器(HIRFL)提供的中能惰性气体离子束(~(20)Ne,122 MeV)作为模拟辐照条件,借助透射电子显微镜,研究了一种低活化的9Cr铁素体/马氏体钢(T92B)组织结构的变化和辐照肿胀.实验结果表明,高温下当材料中晶格原子的撞出损伤和惰性气体原子沉积浓度超过一定限值时,材料内部形成高浓度的空洞,并且空洞肿胀率显著依赖于辐照温度和剂量;在马氏体板条界面及其它晶界处空洞趋于优先形成,并且在晶界交汇处呈加速生长.

  11. Microstructural defects in EUROFER 97 after different neutron irradiation conditions

    Directory of Open Access Journals (Sweden)

    Christian Dethloff

    2016-12-01

    Full Text Available Characterization of irradiation induced microstructural evolution is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM is used to determine the defect microstructure after different neutron irradiation conditions. In particular dislocation loops, voids and precipitates are analyzed concerning defect nature, density and size distribution after irradiation to 15 dpa at 300 °C in the mixed spectrum High Flux Reactor (HFR. New results are combined with previously obtained data from irradiation in the fast spectrum BOR-60 reactor (15 and 32 dpa, 330 °C, which allows for assessment of dose and dose rate effects on the aforementioned irradiation induced defects and microstructural characteristics.

  12. Evaluation of Ion Irradiation Behavior of ODS Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin Sung; Kim, Min Chul; Hong, Jun Hwa; Han, Chang Hee; Chang, Young Mun; Bae, Chang Soo; Bae, Yoon Young; Chang, Moon Hee

    2006-08-15

    FM steel (Grade 92) and ODS alloy(MA956) specimens were ion irradiated with 122 MeV Ne ions. Irradiation temperatures were about 450 and 550 .deg. C and the peak dose was 1, 5, and 10 dpa. Cross-sectional TEM samples were prepared by the electrolytic Ni-plating after pre-treatment of the irradiated specimens. Irradiation cavities in FM steel and ODS alloy specimens were not much different in size; about 20 nm in diameter in both specimens irradiated at around 450 .deg. C. However, the size distribution of cavities in FM steel specimens was broader than that in ODS alloy specimen, indicating that the cavity growth probably via coalescence). It was noticeable that the location and the preferential growth of the cavities in FM steel specimens: cavities on the PAGB (prior austenite grain boundary) was significantly larger than those within the grains. This could be an important issue for the mechanical properties, especially high temperature creep, fracture toughness, and so on. The dependency of the dose threshold and swelling on the ratio of the inert gas concentration/dpa was analysed for the various irradiation source, including He, Ne, Fe/He, and fast neutron, and the empirical correlation was established.

  13. Market challenges for steel

    Energy Technology Data Exchange (ETDEWEB)

    Lauprecht, W.E.; Bulla, W.

    1981-11-01

    Country-wise division of generation of high-alloyed steels, stainless steel and alloyed steel in Western Europe/the Western World. Review of expanding markets for alloyed steels on sectors like-oil field pipes, offshore structure gas- and oil transport in pipelines, coal conversion, nuclear energy, condenser tubes, solar energy, car industry, environmental protection and chemistry.

  14. Instrumented charpy impact tests of austenitic and ferritic steels

    Science.gov (United States)

    Yoshida, H.; Miyata, K.; Hayashi, Y.; Narui, M.; Kayano, H.

    1985-08-01

    The instrumented Charpy impact test was applied to commercial Mn-steel and ferritic steels before and after JMTR irradiation ( 6.5 × 10 22 n/m 2). The load-deflection curves show typical characteristics of the fracture properties of the specimens; i.e. linear elastic behaviour for the brittle fracture and elastic-plastic behaviour for the ductile fracture. The fracture deflection and the absorption energy (fracture energy) dropped rapidly at the temperature of ductile to brittle transition. The ductile-brittle transition temperatures (DBTTs) showed shifts of about 30 and 40 K due to the irradiation for 9Cr-1Mo and 9Cr-2Mo steels, respectively. In Mn-steel the transition from ductile to brittle did not appear at temperatures higher than 77 K. The lateral expansions measured from the scanning electron micrographs show good correspondence to the above results.

  15. Plasma nitriding of steels

    CERN Document Server

    Aghajani, Hossein

    2017-01-01

    This book focuses on the effect of plasma nitriding on the properties of steels. Parameters of different grades of steels are considered, such as structural and constructional steels, stainless steels and tools steels. The reader will find within the text an introduction to nitriding treatment, the basis of plasma and its roll in nitriding. The authors also address the advantages and disadvantages of plasma nitriding in comparison with other nitriding methods. .

  16. Precipitation of the Carbides $M_{23}C_{6}$ under the Irradiation by High Energy Heavy Ions

    CERN Document Server

    Hofmann, A; Semina, V K; Kochanski, T

    2000-01-01

    Carbide M_{23}C_{6} precipitation process in chromium-nickel steels 12H18N9ô and 00H17N14í2 irradiated with high energy heavy Ar^{+6} ions at 625^{o}ó has been studied. It was found that ion irradiation accelerates carbide M_{23}C_{6} precipitation in comparison to thermal annealing. It was shown that composition of carbides formed by irradiation in 00H17N14í2 steel formed under irradiation differs from composition of carbides precipitated during thermal ageing.

  17. Ultrahigh carbon steels, Damascus steels, and superplasticity

    Energy Technology Data Exchange (ETDEWEB)

    Sherby, O.D. [Stanford Univ., CA (United States). Dept. of Materials Science and Engineering; Wadsworth, J. [Lawrence Livermore National Lab., CA (United States)

    1997-04-01

    The processing properties of ultrahigh carbon steels (UHCSs) have been studied at Stanford University over the past twenty years. These studies have shown that such steels (1 to 2.1% C) can be made superplastic at elevated temperature and can have remarkable mechanical properties at room temperature. It was the investigation of these UHCSs that eventually brought us to study the myths, magic, and metallurgy of ancient Damascus steels, which in fact, were also ultrahigh carbon steels. These steels were made in India as castings, known as wootz, possibly as far back as the time of Alexander the Great. The best swords are believed to have been forged in Persia from Indian wootz. This paper centers on recent work on superplastic UHCSs and on their relation to Damascus steels. 32 refs., 6 figs.

  18. Irradiation creep of various ferritic alloys irradiated {approximately}400 C in the PFR and FFTF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Toloczko, M.B. [Washington State Univ., WA (United States); Garner, F.A. [Pacific Northwest National Lab., Richland, WA (United States); Eiholzer, C.R. [Westinghouse Hanford Co., WA (United States)

    1998-03-01

    Three ferritic alloys were irradiated in two fast reactors to doses of 50 dpa or more at temperatures near 400 C. One martensitic alloy, HT9, was irradiated in both the FFTF and PFR reactors. PFR is the Prototype Fast Reactor in Dourneay, Scotland, and FFTF is the Fast Flux Test Facility in Richland, WA. D57 is a developmental alloy that was irradiated in PFR only, and MA957 is a Y{sub 2}O{sub 3} dispersion-hardened ferritic alloy that was irradiated only in FFTF. These alloys exhibited little or no void swelling at {approximately}400 C. Depending on the alloy starting condition, these steels develop a variety of non-creep strains early in the irradiation that are associated with phase changes. Each of these alloys creeps at a rate that is significantly lower than that of austenitic steels irradiated in the same experiments. The creep compliance for ferritic alloys in general appears to be {approximately}0.5 {times} 10{sup {minus}6} MPa{sup {minus}1} dpa{sup {minus}1}, independent of both composition and starting state. The addition of Y{sub 2}O{sub 3} as a dispersoid does not appear to change the creep behavior.

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

    Science.gov (United States)

    Haušild, Petr; Materna, Aleš; Kytka, Miloš

    2015-04-01

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

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

    Science.gov (United States)

    Dethloff, Christian; Gaganidze, Ermile; Aktaa, Jarir

    2014-11-01

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

  1. Evaluation of stainless steels for their resistance to intergranular corrosion

    Science.gov (United States)

    Korostelev, A. B.; Abramov, V. Ya.; Belous, V. N.

    1996-10-01

    Austenitic stainless steels are being considered as structural materials for first wall/blanket systems in the International Thermonuclear Reactor (ITER). The uniform corrosion of stainless steels in water is well known and is not a critical issue limiting its application for the ITER design. The sensitivity of austenitic steels to intergranular corrosion (IGC) can be estimated rather accurately by means of calculation methods, considering structure and chemical composition of steel. There is a maximum permissible carbon content level, at which sensitization of stainless steel is eliminated: K = Cr eff - αC eff, where α-thermodynamic coefficient, Cr eff-effective chromium content (regarding molybdenum influence) and C eff-effective carbon content (taking into account nickel and stabilizing elements). Corrosion tests for 16Cr11Ni3MoTi, 316L and 316LN steel specimens, irradiated up to 2 × 10 22 n/cm 2 fluence have proved the effectiveness of this calculation technique for determination of austenitic steels tendency to IGC. This method is directly applicable in austenitic stainless steel production and enables one to exclude complicated experiments on determination of stainless steel susceptibility to IGC.

  2. Evaluation of stainless steels for their resistance to intergranular corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Korostelev, A.B. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Abramov, V.Ya. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Belous, V.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation)

    1996-10-01

    Austenitic stainless steels are being considered as structural materials for first wall/blanket systems in the international thermonuclear reactor (ITER). The uniform corrosion of stainless steels in water is well known and is not a critical issue limiting its application for the ITER design. The sensitivity of austenitic steels to intergranular corrosion (IGC) can be estimated rather accurately by means of calculation methods, considering structure and chemical composition of steel. There is a maximum permissible carbon content level, at which sensitization of stainless steel is eliminated: K=Cr{sub eff}-{alpha}C{sub eff}, where {alpha}-thermodynamic coefficient, Cr{sub eff}-effective chromium content (regarding molybdenum influence) and C{sub eff}-effective carbon content (taking into account nickel and stabilizing elements). Corrosion tests for 16Cr11Ni3MoTi, 316L and 316LN steel specimens, irradiated up to 2 x 10{sup 22} n/cm{sup 2} fluence have proved the effectiveness of this calculation technique for determination of austenitic steels tendency to IGC. This method is directly applicable in austenitic stainless steel production and enables one to exclude complicated experiments on determination of stainless steel susceptibility to IGC. (orig.).

  3. The effects of low dose rate irradiation and thermal aging on reactor structural alloys

    Science.gov (United States)

    Allen, T. R.; Trybus, C. L.; Cole, J. I.

    As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10 -8 dpa/s) irradiation at 380-410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.

  4. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

    Progress is reported in eight sections: analysis and evaluation studies, test matrices and test methods development, Path A Alloy Development (austenitic stainless steels), Path C Alloy Development (Ti and V alloys), Path D Alloy Development (Fe alloys), Path E Alloy Development (ferritic steels), irradiation experiments and materials inventory, and materials compatibility and hydrogen permeation studies. (DLC)

  5. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-04-01

    Progress is reported in eight sections: analysis and evaluation studies, test matrices and test methods development, Path A Alloy Development (austenitic stainless steels), Path C Alloy Development (Ti and V alloys), Path D Alloy Development (Fe alloys), Path E Alloy Development (ferritic steels), irradiation experiments and materials inventory, and materials compatibility and hydrogen permeation studies. (DLC)

  6. Alloy development for irradiation performance. Quarterly progress report for period ending December 31, 1979

    Energy Technology Data Exchange (ETDEWEB)

    Ashdown, B.G. (comp.)

    1980-04-01

    Progress is reported concerning preparation of a materials handbook for fusion, creep-fatigue of first-wall structural materials, test results on miniature compact tension fracture toughness specimens, austenitic stainless steels, Fe-Ni-Cr alloys, iron-base alloys with long-range crystal structure, ferritic steels, irradiation experiments, corrosion testing, and hydrogen permeation studies. (FS)

  7. Low temperature embrittlement behaviour of different ferritic-martensitic alloys for fusion applications

    Science.gov (United States)

    Rieth, M.; Dafferner, B.

    1996-10-01

    In the last few years a lot of different low activation CrWVTa steels have been developed world-wide. Without irradiation some of these alloys show clearly a better low temperature embrittlement behaviour than commercial CrNiMoV(Nb) alloys. Within the MANITU project a study was carried out to compare, prior to the irradiation program, the embrittlement behaviour of different alloys in the unirradiated condition performing instrumented Charpy impact bending tests with sub-size specimens. The low activation materials (LAM) considered were different OPTIFER alloys (Forschungszentrum Karlsruhe), F82H (JAERI), 9Cr2WVTa (ORNL), and GA3X (PNL). The modified commercial 10-11% CrNiMoVNb steels were MANET and OPTIMAR. A meaningful comparison between these alloys could be drawn, since the specimens of all materials were manufactured and tested under the same conditions.

  8. Reinforcing the Steel Sector

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    By pushing forward mergers between steel-makers, China gears up to consolidate the large but fragmented industryIn a government effort to consolidate the crowded steel industry and position it for fierce global competition, the state-

  9. Phase Stability under Irradiation of Precipitates and Solid Solutions in Model ALloys and in ODS Alloys Relevant for Gen IV

    Energy Technology Data Exchange (ETDEWEB)

    Arthur T. Motta; Robert C. Birtcher

    2007-10-17

    The overall objective of this program is to investigate the irradiation-altered phase stability of oxide precipitates in ODS steels and of model alloy solid solutions of associated systems. This information can be used to determine whether the favorable mechanical propertiies of these steels are maintained under irradiation, thus addressing one of the main materials research issues for this class of steels as identified by the GenIV working groups. The research program will also create fundamental understanding of the irradiation precipitation/dissolution problem by studying a "model" system in which the variables can be controlled and their effects understood individually.

  10. 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.

  11. Development of ODS ferritic-martensitic steels for application to high temperature and irradiation environment; Developpement d'une nouvelle nuance martensitique ODS pour utilisation sous rayonnement a haute temperature

    Energy Technology Data Exchange (ETDEWEB)

    Lambard, V

    2000-07-01

    Iron oxide dispersion strengthened alloys are candidate for nuclear fuel cladding. Therefore, it is crucial to control their microstructure in order to optimise their mechanical properties at temperatures up to 700 deg C. The industrial candidates, ODS ferritic alloys, present an anisotropic microstructure which induces a weakening of mechanical properties in transversal direction as well as the precipitation of brittle phases under thermal aging and irradiation. For this purpose, we tried to develop a material with isotropic properties. We studied several 9Cr-1Mo ferritic/martensitic alloys, strengthened or not by oxide dispersion. The mechanical alloying was performed by attribution and powders were consolidated by hot extrusion. In this work, different metallurgical characterisation techniques and modelling were used to optimise a new martensitic ODS alloy. Microstructural and chemical characterization of matrix has been done. The effect of austenitizing and isochronal tempering treatments on microstructure and hardness has been studied. Oxide distribution, size and chemical composition have been studied before and after high temperature thermal treatment. The study of phase transformation upon heating has permitted the extrapolation to the equilibrium temperature formation of austenite. Phase transformation diagrams upon cooling have been determined and the transformation kinetics have been linked to austenite grain size by a simple relation. Fine grain size is unfavourable for the targeted application, so a particular thermal treatment inducing a coarser grain structure has been developed. Finally, tensile properties have been determined for the different microstructures. (author)

  12. Positron Annihilation Studies of VVER Type Reactor Steels

    OpenAIRE

    Brauer, G.

    1995-01-01

    A summary of recent positron annihilation work on Russian VVER type reactor steels is presented. Thereby, special attention is paid to the outline of basic processes that might help to understand the positron behaviour in this class of industrial material. The idea of positron trapping by irradiation-induced precipitates, which are probably carbides, is discussed in detail.

  13. The steel scrap age.

    Science.gov (United States)

    Pauliuk, Stefan; Milford, Rachel L; Müller, Daniel B; Allwood, Julian M

    2013-04-02

    Steel production accounts for 25% of industrial carbon emissions. Long-term forecasts of steel demand and scrap supply are needed to develop strategies for how the steel industry could respond to industrialization and urbanization in the developing world while simultaneously reducing its environmental impact, and in particular, its carbon footprint. We developed a dynamic stock model to estimate future final demand for steel and the available scrap for 10 world regions. Based on evidence from developed countries, we assumed that per capita in-use stocks will saturate eventually. We determined the response of the entire steel cycle to stock saturation, in particular the future split between primary and secondary steel production. During the 21st century, steel demand may peak in the developed world, China, the Middle East, Latin America, and India. As China completes its industrialization, global primary steel production may peak between 2020 and 2030 and decline thereafter. We developed a capacity model to show how extensive trade of finished steel could prolong the lifetime of the Chinese steelmaking assets. Secondary steel production will more than double by 2050, and it may surpass primary production between 2050 and 2060: the late 21st century can become the steel scrap age.

  14. Thermo-structural analysis and design consideration of the replaceable backwall in IFMIF liquid lithium target

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, H. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195 (Japan)]. E-mail: nakamura.hiroo@jaea.go.jp; Ida, M. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195 (Japan); Chida, T. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195 (Japan); Shiba, K. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195 (Japan); Shimizu, K. [Mitsubishi Heavy Industry, Hyogo 652-8585 (Japan); Sugimoto, M. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki 319-1195 (Japan)

    2007-08-01

    The IFMIF is an accelerator-based intense neutron source for testing candidate materials for fusion reactors. Intense neutrons are emitted inside the Li flow through a backwall. The backwall made of 316L stainless steel or RAFM is attached to the target assembly with a lip seal welded by a YAG laser. Since the backwall is operating under a severe neutron irradiation of 50 dpa/year and a maximum nuclear heating rate of 25 W/cm{sup 3}, thermo-structural design is one of critical issues in a target design. Thermal stress was calculated using the ABAQUS code. As a permissible stress, yield strength at 300 {sup o}C was used. In the case of the 316 stainless steel backwall, the maximum thermal stress was more than the permissible stress (164 MPa). On the other hand, in case of the F82H backwall, a maximum thermal stress was below the permissible stress (455 MPa). Therefore, F82H is recommended as the backwall material.

  15. Influence of grain size on radiation effects in a low carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Alsabbagh, Ahmad, E-mail: ahalsabb@ncsu.edu [Department of Nuclear Engineering, North Carolina State University (United States); Valiev, Ruslan Z. [Institute of Physics of Advanced Materials, Ufa State Aviation Technical University (Russian Federation); Murty, K.L. [Department of Nuclear Engineering, North Carolina State University (United States)

    2013-11-15

    Ultra-fine grain (UFG) metals with a relatively large volume of interfaces are expected to be more radiation resistant than conventional metals; grain boundaries act as unsaturable sinks for neutron irradiation induced defects. Effects of neutron irradiation on conventional and ultra-fine grain structured carbon steel are studied using the PULSTAR reactor at NC State University to relatively low fluence (∼1.15 × 10{sup −3} dpa). The low dose irradiation of ultrafine grained carbon steel revealed minute radiation effects in contrast to the observed radiation hardening and reduction of ductility in its conventional grained counterpart.

  16. Development of advanced low alloy steel for nuclear RPV

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. C.; Shin, K. S.; Lee, S. H.; Lee, B. J. [Seoul National Univ., Seoul (Korea)

    2000-04-01

    Low carbon low alloy steels are used in nuclear power plants as pressure vessel, steam generator, etc. Nuclear pressure vessel material requires good combination of strength/ toughness, good weldability and high resistance to neutron irradiation and corrosion fatigue. For SA508III steels, most widely used in the production of nuclear power plant, attaining toughness is more difficult than strength. When taking into account the loss of toughness due to neutron irradiation, attaining as low transition temperature as possible prior to operation is a critical task in the production of nuclear pressure vessels. In the present study, we investigated detrimental microstructural features of SA508III steels to toughness, then alloy design directions to achieve improved mechanical properties were devised. The next step of alloy design was determined based on phase equilibrium thermodynamics and obtained results. Low carbon low alloy steels having low transition temperatures with enough strength and hardenability were developed. Microstructure and mechanical properties of HAZ of SA508III steels and alloy designed steels were investigated. 22 refs., 147 figs., 38 tabs. (Author)

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

    Science.gov (United States)

    Kenik, E. A.; Busby, J. T.; Gussev, M. N.; Maziasz, P. J.; Hoelzer, D. T.; Rowcliffe, A. F.; Vitek, J. M.

    2017-01-01

    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.

  18. Effects of manufacturing process on impact properties and microstructures of ODS steels

    Energy Technology Data Exchange (ETDEWEB)

    Tanno, Takashi, E-mail: tanno.takashi@jaea.go.jp; Ohtsuka, Satoshi; Yano, Yasuhide; Kaito, Takeji; Tanaka, Kenya

    2014-12-15

    Oxide disp