Study of peculiarities of hydrogen isotopes mixture permeation through low activated steel F82H

International Nuclear Information System (INIS)

Kenzhin, Ye.A.; Tazhibayeva, I.L; Kulsartov, T.V.; Shestakov, V.P.; Chikhray, Ye.V.; Afanasev, S.E.; Zheldak, Yu.L.

2003-01-01

Full text: The problem of diffusion tritium leakage through blanket materials of future fusion device makes some constructive difficulties concerned with protection of personnel and environment and also with losses of tritium, which is planned to be used in the same device. One of the little-studied problems in the tritium leakage process in Fusion Power Plant is that in fact tritium will penetrate through materials while other hydrogen isotopes are present. These are deuterium and hydrogen which always are present in metals. Therefore, for evaluation of tritium leakage in future Fusion Power Plant under such conditions it is necessary to have experimental data about permeation of these hydrogen isotopes through the structure materials.One of proposed structure materials of fusion reactor blanket is low activated steel F82H. The experiment results on evaluation of .hydrogen, deuterium and its mixture interaction parameters with steel F82H are shown in this work. The tests were carried out within temperature range 273-973 K under inlet hydrogen pressure of 100-2000 Pa. Diffusivity, deuterium and hydrogen permeation constants for low activated steel F82H was determined from experiment results. Those experimental results were used for created phenomenology model which describes hydrogen isotope penetration through tube sample from hydrogen isotopes mixture. That model was used so determining the ratios of desorption rates (D-D, D-H, H-H) on outlet side of sample. Using of so obtained results, we can correctly evaluate, the titanium leakage from blanket of fusion machine which will be constructed using low activated steel F82H

Microstructural evolutions of friction stir welded F82H steel for fusion applications

Energy Technology Data Exchange (ETDEWEB)

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.

Small specimen test technology of fracture toughness in structural material F82H steel for fusion nuclear reactors

International Nuclear Information System (INIS)

Wakai, Eiichi; Ohtsuka, Hideo; Jitsukawa, Shiro; Matsukawa, Shingo; Ando, Masami

2006-03-01

Small specimen test technology (SSTT) has been developed to investigate mechanical properties of nuclear materials. SSTT has been driven by limited availability of effective irradiation volumes in test reactors and accelerator-based neutron and charged particle sources, and it is very useful for the reduction of waste materials produced in nuclear engineering. In this study new bend test machines have been developed to obtain fracture behaviors of F82H steel for very small bend specimens of pre-cracked t/2-1/3CVN (Charpy V-notch) with 20 mm-length and DFMB (deformation and fracture mini bend specimen) with 9 mm-length and disk compact tension of 0.18DCT type, and fracture behaviors were examined to evaluate DBTT (ductile-brittle transition temperature) at temperature from -180 to 25degC. The effect of specimen size on DBTT of F82H steel was also examined by using Charpy type specimens such as 1/2t-CVN, 1/3CVN and t/2-1/3CVN. In this paper, it also provides the information of the specimens irradiated at 250degC and 350degC to about 2 dpa in the capsule of 04M-67A and 04M-68A of JMTR experiments. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Microstructural development in multi-pass TIG welded F82H steels under dual-ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Ogiwara, H.; Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Kishimoto, H.; Tomohiro, M.; Kohyama, A. [Kyoto Univ., Institute of Advanced Energy (Japan)

2007-07-01

Full text of publication follows: Reduced-activation ferritic/martensitic (RAFM) steels are the first candidate materials for blanket structural component. On fabricating blanket components, various joining technologies will be required, and it could be anticipated that some parts of the weldments will suffer from high-dose 14 MeV neutron irradiation. For example, the current Japanese water cooled test blanket module design has the weldment of the first wall and side wall at 25 mm from the surface. This will not be the case for TIER-TBM, but could be the critical issue if the same design is applied for DEMO blanket system. In this study, the effects of displacement damage and helium production on mechanical properties and microstructures of the multi-pass tungsten inert gas (TIG) welding were investigated. The welded joint used in this study is a multi-pass TIG weldment on 15-mm thick plates of F82H-IEA. The post weld heat treatment was carried out at 720 deg. C for 1 h. To obtain systematic and accurate information of the microstructural response under fusion environment, dual-ion irradiation was performed. 15 mm x 20 mm pieces which cover the whole TIG weldment were irradiated at 470 deg. C up to 20 dpa using 6.4 MeV Fe{sup +3} and/or energy-degraded 1.0 MeV He{sup +}. The damage rate was 1.0 x 10{sup -3} dpa/s, and the helium injection rate was 15 x 10{sup -3} appm He/s. After the irradiations, thin film samples were made from various regions of the weldment by focused ion beam (FIB) processor, and transmission electron microscopy (TEM) observation and hardness tests were performed. Microstructure and Vickers hardness profiles across base metal, heat affected zone (HAZ) and fusion zone (FZ) were examined before irradiation experiments. The hardness measurements revealed that the maximum hardness was observed at the last pass region of FZ, and the softest region was in the middle of the FZ and HAZ region near the transformation line. In the microstructure study

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

International Nuclear Information System (INIS)

Serra, E.

1997-01-01

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.)

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.

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.

Diffusion bonding of reduced activation ferritic steel F82H for demo blanket application

International Nuclear Information System (INIS)

Kurasawa, T.; Tamura, M.

1996-01-01

A reduced activation ferritic steel, a grade F82H developed by JAERI, is a promising candidate structural material for the blanket and the first wall of DEMO reactors. In the present study, diffusion bonding of F82H has been investigated to develop the fabrication procedures of the blanket box and the first wall panel with cooling channels embedded by F82H. The parameters examined are the bonding temperature (810-1050 C), bonding pressure (2-10 MPa) and roughness of the bonding surface (0.5-12.8 μR max ), and metallurgical examination and mechanical tests of the diffusion bonded joints have been conducted. From the tests, sufficient bonding was obtained under the temperatures of 840-1 050 C (compressive stress of 3-12 MPa), and it was found that heat treatment following diffusion bonding is essential to obtain the mechanical properties similar to that of the base metal. (orig.)

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)

In-pile Tritium Permeation through F82H Steel with and without a Ceramic Coating of Cr2O3-SiO2 Including CrPO4

International Nuclear Information System (INIS)

Nakamichi, M.; Hayashi, K.; Kulsartov, T.V.; Afanasyev, S.E.; Shestakov, V.P.; Chikhray, Y.V.; Kenzhin, E.A.; Kolbaenkov, A.N.

2006-01-01

Development of coating on blanket structural materials with significant reduction capability of tritium permeation is highly required in order to realize a reasonable design of a tritium recovery and processing system of demonstration (DEMO) fusion reactors. An effective coating has been developed in Japan Atomic Energy Agency (JAEA) using a ceramic material of Cr 2 O 3 -SiO 2 including CrPO 4 . In previous out-of-pile deuterium permeation experiments at 600 o C [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701], a significant permeation reduction factor (PFR) of about 300 was obtained for the coating on the inner-side surface of tubular diffusion cells made by ferritic steel (F82H). In the present study, in-pile experiments on tritium permeation were conducted for F82H steel with and without the same coating, using a testing reactor IGV-1M in Kazakhstan. The tritium source used was liquid lithium-lead eutectics, Pb17Li, which was poured into a space around a tubular diffusion cell (specimen) of F82H steel with or without the coating on the inner side the cell. The irradiation time was about 4 hours, which corresponds to a fast-neuron fluence of about 2x10 21 m -2 (E > 1.1 MeV). The permeation reduction factor (PRF) was obtained by comparison of kinetics curves of tritium permeation through the diffusion cell of F82H steel with and without the coating. The PRFs at 600 and 500 o C were 292 and 30, respectively. These values are close to corresponding PRF values of 307 and 45, which had been obtained at 600 and 500 o C, respectively, in the previous out-of-pile experiments [T.V. Kulsartov et al., Fusion Eng. Des. 81 (2006) 701]. (author)

Positron annihilation lifetime measurements of vanadium alloy and F82H irradiated with fission and fusion neutrons

International Nuclear Information System (INIS)

Sato, K.; Inoue, K.; Yoshiie, T.; Xu, Q.; Wakai, E.; Kutsukake, C.; Ochiai, K.

2009-01-01

V-4Cr-4Ti, F82H, Ni and Cu were irradiated with fission and fusion neutrons at room temperature and 473 K. Defect structures were analyzed and compared using positron annihilation lifetime measurement, and microstructural evolution was discussed. The mean lifetime of positrons (the total amount of residual defects) increased with the irradiation dose. The effect of cascade impact was detected in Ni at room temperature. The size and the number of vacancy clusters were not affected by the displacement rate in the fission neutron irradiation at 473 K for the metals studied. The vacancy clusters were not formed in V-4Cr-4Ti irradiated at 473 K in the range of 10 -6 -10 -3 dpa. In F82H irradiated at 473 K, the defect evolution was prevented by pre-existing defects. The mean lifetime of positrons in fission neutron irradiation was longer than that in fusion neutron irradiation in V-4Cr-4Ti at 473 K. It was interpreted that more closely situated subcascades were formed in the fusion neutron irradiation and subcascades interacted with each other, and consequently the vacancy clusters did not grow larger.

Oxide formation and precipitation behaviors on interface of F82H steel joints during HIPing and hot pressing

International Nuclear Information System (INIS)

Kishimoto, H.; Ono, T.; Sakasegawa, H.; Tanigawa, H.; Ando, M.; Shibayama, T.; Kohno, Y.; Kohyama, A.

2013-01-01

Joining technologies for F82H steels are important issues for the development of fusion energy. The hot isostatic pressing (HIP) method is appropriate for consolidating the first wall of the blanket because of the flexibility of the shape of HIPed products. The HIP method is planned for fabricating a complex-shaped first wall component with built-in cooling channels; thus, accumulation of studies of microstructural and mechanical property changes, especially lower toughness of the HIPed joints, is essential. Present research aims to reveal the microstructural evolution of F82H joints fabricated by the HIP method compared with joints fabricated by hot pressing, focusing on the formation of oxides on the interface. F82H joints were characterized using 1/3-scale Charpy V-notch impact test, transmission electron microscope, scanning electron microscope, and electron probe analysis to survey the microstructural characteristics of the interface

A microstructural study of F82H-MOD. Plate, powder HIP and EB and TIG weldments

International Nuclear Information System (INIS)

Schuring, E.W.

1998-06-01

F82H is the heat identification number of a ferritic/martensitic steel. The steel has a chemical composition resulting in low activation after neutron irradiation. At ECN a microstructural study was conducted on plates of F82H (73, 15 and 25mm thick), a bar of HIPped F82H and EB and TIG welds in F82H plates. The investigation was aimed to characterise the microstructures for reference in irradiation experiments. The F82H mod. plates have a martensitic matrix with carbide precipitation on the prior austenite grain boundaries and the martensite laths. The carbide precipitation is most pronounced on the prior austenite grain boundaries. The PAG is comparable for the three plate thicknesses investigated. In two plates faint segregation bands were found with a denser carbide precipitation and 8-ferrite. Further Ta, Ti and Zr containing oxides are present. Hardness measurements show a higher overall hardness in the 25mm plate investigated compared to the 7.5 and 15mm plates. The F82H HIP bar shows a homogeneous PAG throughout the diameter. The microstructure shows a zone starting at about 1 mm below the surface with a width of 14mm, with a relatively high concentration of low precipitation density ferritic like fields, which corresponds to a zone with a 15HV5 lower hardness with respect to the rest of the bar. This zone is expected to be related to the HIP process. Further in the microstructure a circular pattern of W and Ta enriched inclusions is found, which mark the original HIP-powder grain boundaries. The EB-welds were manufactured by two manufactures and differ mostly in the heat input, resulting in differences in HAZ-width and the width of the weld cap. The F82H EB welds show comparable microstructures in the weld pool, with a martensitic microstructure free of carbides. All EB-welds contain 8-ferrite in concentrations of 0.1-1%. In the heat affected zone (HAZ), the coarse grained HAZ (CGHAZ) is free of carbide precipitation, whereas the fine grained HAZ (FGHAZ

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.

Effect of He implantation on fracture behavior and microstructural evolution in F82H

Energy Technology Data Exchange (ETDEWEB)

Yabuuchi, Kiyohiro, E-mail: kiyohiro.yabuuchi@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, 6-6-01-2, Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Sato, Kiminori; Nogami, Shuhei; Hasegawa, Akira [Department of Quantum Science and Energy Engineering, Tohoku University, 6-6-01-2, Aramaki-Aza-Aoba, Aobaku, Sendai, Miyagi 980-8579 (Japan); Ando, Masami; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166, Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aomori 039-3212 (Japan)

2014-12-15

Reduced-activation ferritic/martensitic steels (RAFMs) are the primary candidate structural materials for fusion reactor blanket components. He bubbles, which formed under 14 MeV neutron irradiation, is considered to cause some mechanical property changes. In a previous study, Hasegawa et al. investigated the fracture behavior using Charpy impact test of He implanted F82H by 50 MeV α-particles with cyclotron accelerator, and the ductile brittle transition temperature (DBTT) was increased and intergranular fracture (IGF) was observed. However, the cause of the IGF was not shown in the previous study. To clarify the cause of the IGF of the He implanted F82H by 50 MeV α-particles with cyclotron accelerator, the microstructure of the He implanted F82H was investigated. After Charpy impact test at 233 K, the brittle fracture surface of the He implanted specimen was observed by SEM and TEM. By SEM observation, grain boundary surface was clearly observed from the bottom of the notch to a depth of about 400 μm. This area correspond to the He implanted region. On the other hand, at unimplanted region, river pattern was observed and transgranular fracture occurred. TEM observation revealed the He bubbles agglomeration at dislocations, lath boundaries, and grain boundaries, and the coarsening of precipitates on grain boundaries. IGF of the He implanted F82H was caused by both He bubbles and coarsening precipitates.

Scale structure of aluminised F82H-mod. steel after HIP treatment

International Nuclear Information System (INIS)

Glasbrenner, H.; Stein-Fechner, K.; Konys, J.

2000-01-01

Coatings on low activation steels are required in fusion technology in order to reduce the tritium permeation rate through the steel into the cooling water system by a factor of at least 100. Alumina seems to be a promising coating material. However, an appropriate coating system must also have the potential for self-healing since the ceramic alumina scale tends to fail if mechanical stress is applied. A technology is introduced to form a ductile Al enriched surface scale on F82H-mod steel (Fe-7.7% Cr) and on top of it alumina. This technology consists of two main process steps. Hot dip aluminising has been performed at 700 deg. C for 30 s to the steel in order to introduce Al to the surface near zone by an easy way. The very hard intermetallic scale Fe 2 Al 5 which forms during the immersion process gets completely transformed into FeAl 2 , FeAl and α-Fe(Al) phases during a subsequent HIP process step at high pressure at 1040 deg. C and 30 min. The pressure chosen for the HIP experiment was 3000 bar. Compared to a heat treatment without superimposed high pressure pores formation due to the Kirkendall effect could be suppressed successfully. The influence of the high pressure on the heat treatment (1040 deg. C, 30 min) will be discussed in this paper

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-12-15

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Metallurgical Characterization of Reduced Activation Martensitic Steel F-82H Modified

International Nuclear Information System (INIS)

Fernandez, P.; Lapena, J.; Lancha, A.M.; Gomez-Briceno, D.; Schirra, M.

1999-12-01

During 1995-1998 within of research and development programs on reduced ferritic/martensitic steels for fusion, metallurgical characterization of 8Cr-2WVTa steel, denominated F-28H modified, have been carried out. The work has focused on studying the microstructural and mechanical (tensile, creep, low cycle fatigue and charpy) characteristics of as-received state and aged material in the temperature range 300 degree centigrade to 600 degree centigrade for periods up to 5000 h. (Author) 45 refs

Weldability of dissimilar joint between F82H and SUS316L under fiber laser welding

Energy Technology Data Exchange (ETDEWEB)

Serizawa, Hisashi [Joining and Welding Research Institute, Osaka University, 11-1 Mihogaoka, Ibaraki, Osaka 567-0047 (Japan); Mori, Daiki; Shirai, Yuma; Ogiwara, Hiroyuki; Mori, Hiroaki [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

2013-10-15

Highlights: • The microstructure of F82H/SUS316L dissimilar joint can be divided into four regions. • In the case without beam position shift, hardness of WM cannot be reduced by PWHT. • The fiber laser welding would be applicable for constructing the dissimilar joint. -- Abstract: As one of the high beam quality heat sources, 4 kW fiber laser was applied for joining between reduced activation ferritic/martensitic steel, F82H and SUS316L austenitic stainless steel, and the microstructural analyses and Vickers hardness measurements were carried out before and after post-weld heat treatment (PWHT). The microstructure of joint can be divided into four regions which are base metal of F82H, heat affected zone (HAZ) in F82H, weld metal (WM) and base metal of SUS316L. Also, it is revealed that the high-power fiber laser can be employed for constructing butt joint between F82H and SUS316L by applying PWHT and shifting the laser beam position to SUS316L, where the distance between the contact face and beam should be set as a range from radius to diameter of laser beam.

Weldability of dissimilar joint between F82H and SUS316L under fiber laser welding

International Nuclear Information System (INIS)

Serizawa, Hisashi; Mori, Daiki; Shirai, Yuma; Ogiwara, Hiroyuki; Mori, Hiroaki

2013-01-01

Highlights: • The microstructure of F82H/SUS316L dissimilar joint can be divided into four regions. • In the case without beam position shift, hardness of WM cannot be reduced by PWHT. • The fiber laser welding would be applicable for constructing the dissimilar joint. -- Abstract: As one of the high beam quality heat sources, 4 kW fiber laser was applied for joining between reduced activation ferritic/martensitic steel, F82H and SUS316L austenitic stainless steel, and the microstructural analyses and Vickers hardness measurements were carried out before and after post-weld heat treatment (PWHT). The microstructure of joint can be divided into four regions which are base metal of F82H, heat affected zone (HAZ) in F82H, weld metal (WM) and base metal of SUS316L. Also, it is revealed that the high-power fiber laser can be employed for constructing butt joint between F82H and SUS316L by applying PWHT and shifting the laser beam position to SUS316L, where the distance between the contact face and beam should be set as a range from radius to diameter of laser beam

Hydrogen permeation measurement of the reduced activation ferritic steel F82H by the vacuum thermo-balance method

International Nuclear Information System (INIS)

Yoshida, Hajime; Enoeda, Mikio; Abe, Tetsuya; Akiba, Masato

2005-03-01

Hydrogen permeation fluxes of the reduced activation ferritic steel F82H were quantitatively measured by a newly proposed method, vacuum thermo-balance method, for a precise estimation of tritium leakage in a fusion reactor. We prepared sample capsules made of F82H, which enclosed hydrogen gas. The hydrogen in the capsules permeated through the capsule wall, and subsequently desorbed from the capsule surface during isothermal heating. The vacuum thermo-balance method allows simultaneous measurement of the hydrogen permeation flux by two independent methods, namely, the net weight reduction of the sample capsule and exhaust gas analysis. Thus the simultaneous measurements by two independent methods increase the reliability of the permeability measurement. When the gas pressure of enclosed hydrogen was 0.8 atm at the sample temperature of 673 K, the hydrogen permeation flux of F82H obtained by the net weight reduction and the exhaust gas analysis was 0.75x10 18 (H 2 /m 2 s) and 2.2x10 18 (H 2 /m 2 s), respectively. The ratio of the hydrogen permeation fluxes obtained by the net weight reduction to that measured by the exhaust gas analysis was in the range from 1/4 to 1/1 in this experiment. The temperature dependence of the estimated permeation flux was similar in both methods. Taking the uncertainties of both measurements into consideration, both results are supposed to be consistent. The enhancement of hydrogen permeation flux was observed from the sample of which outer surface was mechanically polished. Through the present experiments, it has been demonstrated that the vacuum thermo-balance method is effective for the measurement of hydrogen permeation rate of F82H. (author)

Water corrosion of F82H-modified in simulated irradiation conditions by heat treatment

International Nuclear Information System (INIS)

Lapena, J.; Blazquez, F.

2000-01-01

This paper presents results of testing carried out on F82H in water at 260 deg. C with 2 ppm H 2 and the addition of 0.27 ppm Li in the form of LiOH. Uniform corrosion tests have been carried out on as-received material and on specimens from welded material [TIG and electron beam (EB)]. Stress corrosion cracking (SCC) tests have been carried out in as-received material and in material heat treated to simulate neutron irradiation hardening (1075 deg. C/30' a.c. and 1040 deg. C/30' + 625 deg. C/1 h a.c.) with hardness values of 405 and 270 HV30, respectively. Results for uniform corrosion after 2573 h of testing have shown weight losses of about 60 mg/dm 2 . Compact tension (CT) specimens from the as-received material tested under constant load have not experienced crack growth. However, in the simulated irradiation conditions for a stress intensity factor between 40 and 80 MPa√m, crack growth rates of about 7x10 -8 m/s have been measured

Materials design data for reduced activation martensitic steel type F82H

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A.-A.F. E-mail: tavassoli@cea.fr; Rensman, J.-W.; Schirra, M.; Shiba, K

2002-11-01

This paper presents materials data for design of ITER test blanket modules with the reduced activation ferritic martensitic steel type F82H as structural material. From the physical properties databases, variations of modulus of elasticity, density, thermal conductivity, thermal diffusivity, specific heat, mean and instantaneous linear coefficients of thermal expansion versus temperature are derived. Also reported are Poisson's ratio and magnetic properties. From the tension test results, nominal and minimum stress values of S{sub y} and S{sub u} are derived and used for calculation of allowable primary membrane stress intensity S{sub m}. Likewise, uniform and total elongations, as well as reduction of area data, are used for calculation of minimum and true ductility at rupture values. From the instrumented Charpy impact and fracture toughness test data, ductile to brittle transition temperature, toughness and behavior of material in different fracture modes are evaluated. The effect of specimen size and geometry are discussed but preference is given to standard size specimens. From the fatigue data, total strain range versus number of cycles to failure curves are plotted and used to derive fatigue design curves, using a reduction factor of 2 on strain and a reduction factor of 20 on number of cycles to failure. Cyclic hardening curves are also derived and compared with monotonic hardening curves. From the creep data, time dependent allowable stresses S{sub r} and S{sub t} are calculated. Combination of tension and creep results are used to deduce S{sub mt} and isochronus curves. Finally, irradiated and aged materials data are compared to insure that the safety margins incorporated in unirradiated design limits are not exceeded.

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.

Radiation induced microstructural evolution in ferritic/martensitic steels

International Nuclear Information System (INIS)

Kohno, Y.; Kohyama, A.; Asakura, K.; Gelles, D.S.

1993-01-01

R and D of ferritic/martensitic steels as structural materials for fusion reactor is one of the most important issues of fusion technology. The efforts to characterize microstructural evolution under irradiation in the conventional Fe-Cr-Mo steels as well as newly developed Fe-Cr-Mn or Fe-Cr-W low activation ferritic/ martensitic steels have been continued. This paper provides some of the recent results of heavy irradiation effects on the microstructural evolution of ferritic/martensitic steels neutron irradiated in the FFTF/MOTA (Fast Flux Test Facility/Materials Open Test Assembly). Materials examined are Fe-10Cr-2Mo dual phase steel (JFMS: Japanese Ferritic/Martensitic Steel), Fe-12Cr-XMn-1Mo manganese stabilized martensitic steels and Fe-8Cr-2W Tungsten stabilized low activation martensitic steel (F82H). JFMS showed excellent void swelling resistance similar to 12Cr martensitic steel such as HT-9, while the manganese stabilized steels and F82H showed less void swelling resistance with small amount of void swelling at 640-700 K (F82H: 0.14% at 678 K). As for irradiation response of precipitate behavior, significant formation of intermetallic χ phase was observed in the manganese stabilized steels along grain boundaries which is though to cause mechanical property degradation. On the other hand, precipitates identified were the same type as those in unirradiated condition in F82H with no recognition of irradiation induced precipitates, which suggested satisfactory mechanical properties of F82H after the irradiation. (author)

Basic characterization of low-activated F82H-mod steel as compared to OPTIFER

International Nuclear Information System (INIS)

Schirra, M.; Adelhelm, C.; Graf, P.; Heger, S.; Kempe, H.; Zimmermann, H.

1997-12-01

Low-activated F82H-model steel plate material was made available to the European laboratories by the Japanese side. Basic characterization work is being performed by FZK/IMF. In the present report, the results obtained shall be compared with the European OPTIFER types. The chemical composition and the state as delivered are controlled. In addition, transformation, hardness and tempering behaviors shall be described. Tensile strength and impact bending strength were determined in the ranges of RT to 700 C and -80 C to +80 C, respectively. Particular attention is paid to the creep strength behavior. The current tests cover the range of up to 15000 h at temperatures ranging from 450 to 700 C. It can be noticed that the results measured for some parameters are subject to considerable scattering. As compared to previous tests, the experimental scope therefore had to be increased significantly. (orig.) [de

Behaviour of F82H mod. stainless steel in lead-bismuth under temperature gradient

Science.gov (United States)

Gómez Briceño, D.; Martín Muñoz, F. J.; Soler Crespo, L.; Esteban, F.; Torres, C.

2001-07-01

Austenitic steels can be used in a hybrid system in contact with liquid lead-bismuth eutectic if the region of operating temperatures is not beyond 400°C. For higher temperatures, martensitic steels are recommended. However, at long times, the interaction between the structural material and the eutectic leads to the dissolution of some elements of the steel (Ni, Cr and Fe, mainly) in the liquid metal. In a non-isothermal lead-bismuth loop, the material dissolution takes place at the hot leg of the circuit and, due to the mass transfer, deposition occurs at the cold leg. One of the possible ways to improve the performance of structural materials in lead-bismuth is the creation of an oxide layer. Tests have been performed in a small natural convection loop built of austenitic steel (316L) that has been operating for 3000 h. This loop contains a test area in which several samples of F82Hmod. martensitic steel have been tested at different times. A gas with an oxygen content of 10 ppm was bubbled in the hot area of the circuit during the operation time. The obtained results show that an oxide layer is formed on the samples introduced in the loop at the beginning of the operation and this layer increases with time. However, the samples introduced at different times during the loop operation, are not protected by oxide layers and present material dissolution in some cases.

Behaviour of F82H mod. stainless steel in lead-bismuth under temperature gradient

International Nuclear Information System (INIS)

Gomez Briceno, D.; Martin Munoz, F.J.; Soler Crespo, L.; Esteban, F.; Torres, C.

2001-01-01

Austenitic steels can be used in a hybrid system in contact with liquid lead-bismuth eutectic if the region of operating temperatures is not beyond 400 deg. C. For higher temperatures, martensitic steels are recommended. However, at long times, the interaction between the structural material and the eutectic leads to the dissolution of some elements of the steel (Ni, Cr and Fe, mainly) in the liquid metal. In a non-isothermal lead-bismuth loop, the material dissolution takes place at the hot leg of the circuit and, due to the mass transfer, deposition occurs at the cold leg. One of the possible ways to improve the performance of structural materials in lead-bismuth is the creation of an oxide layer. Tests have been performed in a small natural convection loop built of austenitic steel (316L) that has been operating for 3000 h. This loop contains a test area in which several samples of F82Hmod. martensitic steel have been tested at different times. A gas with an oxygen content of 10 ppm was bubbled in the hot area of the circuit during the operation time. The obtained results show that an oxide layer is formed on the samples introduced in the loop at the beginning of the operation and this layer increases with time. However, the samples introduced at different times during the loop operation, are not protected by oxide layers and present material dissolution in some cases

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

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

International Nuclear Information System (INIS)

Yamamoto, N.; Murase, Y.; Nagakawa, J.

2007-01-01

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 α-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 -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 high

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

Chemical segregation behavior under thermal aging of the low-activation F82H-modified steel

International Nuclear Information System (INIS)

Lapena, J.; Garcia-Mazario, M.; Fernandez, P.; Lancha, A.M.

2000-01-01

In this work, thermal aging of the low-activation F82H-modified steel has been performed at temperatures in the range 300-600 deg. C during periods up to 5000 h. A detailed mechanical and microstructural characterization has been carried out in the aged materials, as well as in the as-received state material for reference. Auger electron spectroscopy (AES) analysis has been performed for these materials to study the microchemistry at the grain boundaries. The results show a decrease of the impact properties after aging at 600 deg. C that has been related to the precipitation of the Laves phase. Auger analyses show chromium enrichment and iron depletion at grain boundaries in all material conditions. In addition, sulphur and tungsten have been observed by this technique at grain boundaries, their presence and distribution being dependent on the material state

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

International Nuclear Information System (INIS)

Spaetig, P.; Bonade, R.; Odette, G.R.; Rensman, J.W.; Campitelli, E.N.; Mueller, P.

2007-01-01

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

Evaluation of toughness degradation by small punch (SP) tests for neutron irradiated structural steels

International Nuclear Information System (INIS)

Misawa, Toshihei; Hamaguchi, Yoshikazu; Kimura, Akihiko; Eto, Motokuni; Suzuki, Masahide; Nakajima, Nobuya.

1992-01-01

The small punch (SP) test as one of the useful small specimen testing technique (SSTT) has been developed to evaluate the fracture toughness, ductile-brittle transition temperature (DBTT) and tensile properties for neutron irradiated structural materials. The SP tests using the miniaturized specimens of φ3 mm TEM disk and 10 mm 2 coupon were performed for six kinds of ferritic steels of F-82, F-82H, HT-9, JFMS, 2.25-1Mo and SQV2A. It was shown that the temperature dependence of SP fracture energies with scatter in miniaturized testing can give reliable information on the DBTT by use of the statistical analysis based on the Weibull distribution. A good correlation between the DBTT of the SP tests and that of the standard CVN test has been obtained for the various nuclear ferritic steels. The SP test was performed for cryogenic austenitic steels as a way of evaluating elastic-plastic fracture toughness, J IC , on the basis of a universal empirical relationship between J IC and SP equivalent fracture strain, ε-bar qf . The SP testing using the neutron irradiated specimens of 2.25Cr-1Mo, F-82, F-82H and HT-9 steels was successfully applied and presented the neutron radiation induced changes on the DBTT, fracture toughness and tensile properties. (author)

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

Science.gov (United States)

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

2012-12-01

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

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.

Self-ion Irradiation Damage of F/M and ODS steels

International Nuclear Information System (INIS)

Kang, Suk Hoon; Chun, Young-Bum; Noh, Sanghoon; Jang, Jinsung; Kim, Tae Kyu

2014-01-01

Oxide dispersion strengthened (ODS) ferritic steels are potential high-temperature materials that are stabilized by dispersed particles at elevated temperatures. These dispersed particles improve the tensile strength and creep rupture strength, they are expected to increase the operation temperature up to approximately 650 .deg. C and also enhance the energy efficiency of the fusion reactor. Some reports described that the nano-clusters are strongly resistant to coarsening by annealing up to 1000 .deg. C, and nanoclusters do not change after ion irradiation up to 0.7 dpa at 300 .deg. C. ODS steels will be inevitably exposed to neutron irradiation condition; the irradiation damages, creep and swelling are always great concern. The dispersed oxide particles are believed to determine the performance of the steel, even the radiation resistance. In this study, F/M and ODS model alloys of Korea Atomic Energy Research Institute (KAERI) were irradiated by Fe 3+ self-ion to emulate the neutron irradiation effect. In this study, Fe 3+ self-ion irradiation is used as means of introducing radiation damage in F/M steel and ODS steel. The ion accelerator named DuET (in Kyoto University, Japan) was used for irradiation of Fe 3+ ion by 6.4 MeV at 300 .deg. C. The maximum damage rate in F/M and ODS steels were estimated roughly 6 dpa. After radiation, point or line defects were dominantly observed in F/M steel, on the other hands, small circular cavities were typically observed in ODS steel. Nanoindentation is a useful tool to determine the irradiationinduced hardness change in the damage layer of ionirradiated iron base alloys

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)

Fracture toughness of the F-82H steel-effect of loading modes, hydrogen, and temperature

International Nuclear Information System (INIS)

Li, H.-X.; Jones, R.H.; Hirth, J.P.; Gelles, D.S.

1996-01-01

The effects of loading mode, hydrogen, and temperature on fracture toughness and tearing modulus were examined for a ferritic/martensitic steel (F-82H). The introduction of a shear load component, mode III, significantly decreased the initiation and propagation resistance of cracks compared to the opening load, mode I, behavior. Mode I crack initiation and propagation exhibited the highest resistance. A minimum resistance occurred when the mode I and mode III loads were nearly equal. The presence of 4 wppm hydrogen decreased the cracking resistance compared to behavior without H regardless of the loading mode. The minimum mixed-mode fracture toughness with the presence of hydrogen was about 30% of the hydrogen-free mode I fracture toughness. The mixed-mode toughness exhibited a lesser sensitivity to temperature than the mode I toughness. The J IC value was 284 kJ/m 2 at room temperature, but only 60 kJ/m 2 at -55 C and 30 kJ/m 2 at -90 C. The ductile to brittle transition temperature (DBTT) was apparently higher than -55 C. (orig.)

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.

Effect of laser beam position on mechanical properties of F82H/SUS316L butt-joint welded by fiber laser

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); Mori, Daiki; Ogiwara, Hiroyuki; Mori, Hiroaki [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

2014-10-15

Highlights: • The micro hardness of weld metal in F82H/SUS316L joint partially decreases after PWHT by shifting beam position to SUS316L. • Charpy impact energy of F82H/SUS316L joint obviously increases after PWHT due to the release of residual stress. • The tensile strength of weld metal in F82H/SUS316L joint is higher than that of SUS316L. • The fiber laser welding seems to be one of the most candidate methods to join between F82H and SUS316L pipes practically. - Abstract: A dissimilar butt-joint between reduced activation ferritic/martensitic steel F82H and SUS316L austenitic stainless steel was made by 4 kW fiber laser and the influence of laser beam position on its mechanical properties before and after post-weld heat treatment (PWHT) was examined at room temperature. From the nano-indentation measurements and the microstructural observations, it is found that the micro hardness of weld metal partially decreases after PWHT by shifting beam position to SUS316L because its phase seems to move from only the martensitic phase to the mixture of austenitic and martensitic phases. In addition, Charpy impact test suggests that the impact energy slightly increases by shifting beam position before PWHT and obviously increases after PWHT due to the release of residual stress. Moreover, the tensile test indicates that the tensile strength of weld metal is higher than that of SUS316L and the fracture occurs at the base metal of SUS316L regardless of laser beam position.

Impact properties of reduced activation ferritic/martensitic steel, F82H jointed by hot isostatic pressing

Energy Technology Data Exchange (ETDEWEB)

Ogiwara, H.; Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Ibaraki-ken (Japan); Kohyama, A. [Kyoto Univ., lnstitute of Advanced Energy (Japan)

2007-07-01

Full text of publication follows: Reduced-activation ferritic/martensitic steels are the leading candidate structural material for the blanket system of fusion reactors. The important issue at the current stage is the finalization of a detailed manufacturing specification for ITER test blanket module. Hot isostatic pressing (HIP) process is one of the most important methods to fabricate the first wall with cooling channels. The objective of this paper is to optimize HIP condition to obtain the excellent joints mechanical properties. The materials used were F82H steels. The joint was produced by solid state HIP method. Before HIP treatments, specimens were heated in vacuum condition to out-gas. This treatment was conducted to decrease oxidation on the surfaces. HIP treatments were carried out for 2 h at 1100 deg. C - 140 MPa. The specimens were normalized at 960 deg. C for 0.5 h and tempered at 750 deg. C for 1.5 h. The bonding interface was characterized by scanning electron microscopy. Charpy impact tests and tensile tests were conducted to evaluate the mechanical properties of the HIP joint. Impact tests revealed that there were no significant differences in the ductile-brittle transition temperatures of HIP jointed specimens and base metal specimens, but the upper-shelf energy (USE) of the HIP joint specimens at room temperature was only about 10% of that of the base metal specimens. SEM observations of the fracture surface of HIP joint specimens revealed that a large number of oxides were formed on the HIP joint. This result indicates that oxides formed on the HIP joint are the dominant factor of the impact properties. Based on these results, the pre-HIP treatment conditions had been optimized to reduce the number of oxides, and USE of HIP joint specimens increased to about 50% of that of the base metal. The detailed analyses on the HIP joint microstructure will be reported. (authors)

Hardening and embrittlement mechanisms of reduced activation ferritic/martensitic steels irradiated at 573 K

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Klueh, R.L. [Oak Ridge Noational Laboratory, TN (United States); Hashimoto, N. [Hokkaido Univ., Materials Science and Engineering Div., Graduate School of Engineering, Sapporo (Japan); Sokolov, M. [Oak Ridge National Laboratory, Materials Science and Technology Div., TN (United States)

2007-07-01

Full text of publication follows: It has been reported that reduced-activation ferritic/martensitic steels (RAFMs), such as F82H, ORNL9Cr-2WVTa, and JLF-1, showed a variety of changes in ductile-brittle transition temperature and yield stress after irradiation at 573 K up to 5 dpa, and those differences could not be interpreted solely by the difference of dislocation microstructure induced by irradiation. To investigate the impact of other microstructural feature, i.e. precipitates, the precipitation behavior of F82H, ORNL 9Cr-2WVTa, and JLF-1 was examined. It was revealed that irradiation-induced precipitation and amorphization of precipitates partly occurred and caused the different precipitation on block, packet and prior austenitic grain boundaries. In addition to these phenomena, irradiation-induced nano-size precipitates were also observed in the matrix. It was also revealed that the chemical compositions of precipitates approached the calculated thermal equilibrium state of M{sub 23}C{sub 6} at an irradiation temperature of 573 K. The calculation also suggests the presence of Laves phase at 573 K, which is usually not observed at this temperature, but the ion irradiation on aged F82H with Laves phase suggests that Laves phase becomes amorphous and could not be stable under irradiation at 573 K. This observation indicates the possibility that the irradiation-induced nano-size precipitation could be the consequence of the conflict between precipitation and amorphization of Laves phase. Over all, these observations suggests that the variety of embrittlement and hardening of RAFMs observed at 573 K irradiation up to 5 dpa might be the consequence of the transition phenomena that occur as the microstructure approaches thermal equilibrium during irradiation at 573 K. (authors)

Interface strength measurement of tungsten coatings on F82H substrates

International Nuclear Information System (INIS)

Kim, Hyoungil; El-Awady, Jaafar; Gupta, Vijay; Ghoniem, Nasr; Sharafat, Shahram

2009-01-01

In the current work, hot isostatic pressing is adopted to deposit tungsten coatings on F82H substrates. The interface strength of the W/F82H samples is measured using the Laser Spallation technique and the microstructure is analyzed to determine the strength of the coating. Finally, the failure mechanisms of the hot isostatic pressing versus vacuum plasma spraying tungsten coatings and their different failure strengths are compared. It is concluded that the hot isostatic pressing process ensures a good adhesion for the W/F82H interface while the vacuum plasma spraying process results in relatively lower failure strength for the W-coating itself due to the high porosity in the coating.

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

International Nuclear Information System (INIS)

Tanigawa, H.; Shiba, K.; Ando, M.; Wakai, E.; Jitsukawa, S.; Hirose, T.; Kasada, R.; Kimura, A.; Kohyama, A.; Kohno, Y.; Klueh, R.L.; Sokolov, M.; Stoller, R.; Zinklek, S.; Yamamoto, T.; Odette, G.; Kurtz, R.J.

2007-01-01

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)

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)

Development of HIP bonding procedure and mechanical properties of HIP bonded joints for reduced activation ferritic steel F-82H

International Nuclear Information System (INIS)

Oda, Masahiro; Kurasawa, Toshimasa; Kuroda, Toshimasa; Hatano, Toshihisa; Takatsu, Hideyuki

1997-03-01

Structural materials of blanket components in fusion DEMO reactors will receive a neutron wall load more than 3-5MW/m 2 as well as exposed by surface heat flux more than 0.5MW/m 2 . A reduced activation ferritic steel F-82H has been developed by JAERI in collaboration with NKK from viewpoints of resistance for high temperature and neutron loads and lower radioactivity. This study intends to obtain basic performance of F-82H to establish the fabrication procedure of the first wall and blanket box by using Hot Isostatic Pressing (HIP) bonding. Before HIP bonding tests, effects of heat treatment temperature and surface roughness on mechanical properties of joints were investigated in the heat treatment tests and diffusion bonding tests, respectively. From these results, the optimum HIP bonding conditions and the post heat treatment were selected. Using these conditions, the HIP bonding tests were carried out to evaluate HIP bondability and to obtain mechanical properties of the joints. Sufficient HIP bonding performance was obtained under the temperature of 1040degC, the compressive stress of 150MPa, the holding time of 2h, and the surface roughness ∼μ m. Mechanical properties of HIP bonded joints with these conditions were similar to those of as-received base metal. An oxide formation on the surface to be bonded would need to be avoided for sufficient bonding. The bonding ratio, Charpy impact value and fatigue performance of the joints strongly depended on the HIP conditions, especially temperature, while micro-structure, Vickers hardness and tensile properties had little dependence on the HIP temperature. The surface roughness strongly affected the bonding ratio and would be required to be in the level of a few μ m. In the HIP bonding test of the welded material, the once-melted surface could be jointed by the HIP bonding under the above-mentioned procedure. (J.P.N.)

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

International Nuclear Information System (INIS)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk; Jeong, Jong Ryul; Maeng, Cheol Soo; Lee, Myoung Goo

2016-01-01

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10"-"5 dpa.

Analysis of Low Dose Irradiation Damages in Structural Ferritic/Martensitic Steels by Proton Irradiation and Nanoindentation

Energy Technology Data Exchange (ETDEWEB)

Waseem, Owais A.; Ryu, Ho Jin; Park, Byong Guk [KAIST, Daejeon (Korea, Republic of); Jeong, Jong Ryul [Chungnam University, Daejeon (Korea, Republic of); Maeng, Cheol Soo; Lee, Myoung Goo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

As a result, ferritic-martensitic steels find applications in the in-core and out-of-core components which include ducts, piping, pressure vessel and cladding, etc. Due to ferromagnetism of F/M steel, it has been successfully employed in solenoid type fuel injector. Although the irradiation induced degradation in ferritic martensitic steels is lower as compare to (i) reduced activation steels, (ii) austenitic steels and (iii) martensitic steels, F/M steels are still prone to irradiation induced hardening and void swelling. The irradiation behavior may become more sophisticated due to transmutation and production of helium and hydrogen. The ductile to brittle transition temperature of F/M steels is also expected to increase due to irradiation. These irradiation induced degradations may deteriorate the integrity of F/M components. As a result of these investigations, it has found that the F/M steels experience no irradiation hardening above 400 .deg. C, but below this temperature, up to 350 .deg. C, weak hardening is observed. The irradiation hardening becomes more pronounced below 300 .deg. C. Moreover, the irradiation hardening has also found dependent upon radiation damage. The hardening was found increasing with increasing dose. Due to pronounced irradiation hardening below 300 .deg. C and increasing radiation damage with increasing dose (even at low dpa), it is required to investigate the post irradiation mechanical properties of F/M steel, in order to confirm its usefulness in structural and magnetic components which experience lifetime doses as low as 1x10{sup -5} dpa.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-15

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

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

International Nuclear Information System (INIS)

Tanigawa, H.; Hirose, T.; Shiba, K.; Kasada, R.; Wakai, E.; Serizawa, H.; Kawahito, Y.; Jitsukawa, S.; Kimura, A.; Kohno, Y.; Kohyama, A.; Katayama, S.; Mori, H.; Nishimoto, K.; Klueh, R.L.; Sokolov, M.A.; Stoller, R.E.; Zinkle, S.J.

2008-01-01

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

Effects of helium on ductile brittle transition behavior of reduced activation ferritic steels after high concentration he implantation at high temperature

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, A.; Ejiri, M.; Nogami, S.; Ishiga, M.; Abe, K. [Tohoku Univ., Dept. of Quantum Science and Energy Engr, Sendai (Japan); Kasada, R.; Kimura, A. [Kyoto Univ., Institute of Advanced Energy (Japan); Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Influence of Helium (He) on fracture behavior of reduced activation ferritic/martensitic steels including Oxide Dispersion Strengthening (ODS) steels and F82H were examined. To study the He effects on fracture behavior of these steels after He bubble formation conditions, higher concentration of He implantation at around 550 C were performed and examined the relationship between microstructure evolution and fracture behavior of the steels. The 1.5CVN mini size Charpy specimens were used to evaluate impact test behavior. Reduced activation ferritic ODS steels, 9Cr-ODS and 12Cr-ODS steels were examine. F82H was also examined as reference material. Helium implantation was performed by a cyclotron of Tohoku University with a beam of 50 MeV {alpha}-particles at temperature around 550 C. A tandem-type energy degrader system was used to implant He into the specimen from the irradiated surface to the range of 50 MeV {alpha}-particles, that was about 380 {mu}m in iron. Implanted He concentration were about 1000 appm. Charpy impact test was performed using a instrumented impact test apparatus in Oarai branch of IMR, Tohoku University. Analyses of absorbed energy change and fracture surface were carried out. Vickers hardness test was also carried out on He implanted area of the 1.5CVN specimen to estimate irradiation hardening. Microstructural observation was performed by TEM. In the case of F82H, DBTT increased by the 1000 appm He implantation condition was about 80 C and grain boundary fracture surface was only observed in the He implanted area of all the ruptured specimens in brittle manner. On the other hand, DBTT shift and fracture mode change of He implanted 9Cr-ODS steel was not observed after He implantation. Microstructural observation showed that He bubble formation on the lath boundaries and grain boundaries were significant in F82H, but the bubble segregation on grain boundary in ODS steel was not apparent. The bubble formation

Corrosion of carbon steel and low-alloy steel in diluted seawater containing hydrazine under gamma-rays irradiation

International Nuclear Information System (INIS)

Nakano, Junichi; Yamamoto, Masahiro; Tsukada, Takashi

2014-01-01

Seawater was injected into reactor cores of Units 1, 2, and 3 in the Fukushima Daiichi nuclear power station as an urgent coolant. It is considered that the injected seawater causes corrosion of steels of the reactor pressure vessel and primary containment vessel. To investigate the effects of gamma-rays irradiation on weight loss in carbon steel and low-alloy steel, corrosion tests were performed in diluted seawater at 50°C under gamma-rays irradiation. Specimens were irradiated with dose rates of 4.4 kGy/h and 0.2 kGy/h. To evaluate the effects of hydrazine (N 2 H 4 ) on the reduction of oxygen and hydrogen peroxide, N 2 H 4 was added to the diluted seawater. In the diluted seawater without N 2 H 4 , weight loss in the steels irradiated with 0.2 kGy/h was similar to that in the unirradiated steels, and weight loss in the steels irradiated with 4.4 kGy/h increased to approximate 1.7 times of those in the unirradiated steels. Weight loss in the steels irradiated in the diluted seawater containing N 2 H 4 was similar to that in the diluted seawater without N 2 H 4 . When N 2 was introduced into the gas phase in the flasks during gamma-rays irradiation, weight loss in the steels decreased. (author)

IRRADIATION CREEP AND MECHANICAL PROPERTIES OF TWO FERRITIC-MARTENSITIC STEELS IRRADIATED IN THE BN-350 FAST REACTOR

International Nuclear Information System (INIS)

Porollo, S. I.; Konobeev, Yu V.; Dvoriashin, A. M.; Budylkin, N. I.; Mironova, E. G.; Leontyeva-Smirnova, M. V.; Loltukhovsky, A. G.; Bochvar, A. A.; Garner, Francis A.

2002-01-01

Russian ferritic/martensitic steels EP-450 and EP-823 were irradiated to 20-60 dpa in the BN-350 fast reactor in the form of pressurized creep tubes and small rings used for mechanical property tests. Data derived from these steels serves to enhance our understanding of the general behavior of this class of steels. It appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures less then 420 degrees C, but may be camouflaged somewhat by precipitation-related densification. The irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels, and that the loss of strength at test temperatures above 500 degrees C is a problem generic to all F/M steels. This conclusion is supported by post-irradiation measurement of short-term mechanical properties. At temperatures below 500 degrees C both steels retain their high strength (yield stress 0.2=550-600 MPa), but at higher test temperatures a sharp decrease of strength properties occurs. However, the irradiated steels still retain high post-irradiation ductility at test temperatures in the range of 20-700 degrees C.

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.

The behaviour of ferritic steels under fast neutron irradiation

International Nuclear Information System (INIS)

Erler, J.; Maillard, A.; Brun, G.; Lehmann, J.; Dupouy, J.M.

1979-07-01

Ferritic steels have been irradiated in Rapsodie and Phenix to doses up to 150 dpa F. The swelling and irradiation creep characteristics and the mechanical properties of these materials are reported. (author)

Finite element modeling and experimental study of brittle fracture in tempered martensitic steels for thermonuclear fusion applications

International Nuclear Information System (INIS)

Mueller, P. F.

2009-10-01

The present report studies the brittle fracture in high-chromium reduced activation tempered martensitic steels foreseen as structural materials for thermonuclear fusion reactors. Developing the adequate materials that can withstand the severe irradiation conditions of the burning plasma in a fusion reactor is one of the major challenges to be solved in order to make profit from the great advantages of thermonuclear fusion as an energy source. High-chromium tempered martensitic steels such as F82H and the most advanced version Eurofer97 are among the main candidate materials for structural applications in future fusion power plants due to low irradiation-induced swelling, good mechanical and thermal properties, and reasonably fast radioactive decay. Drawback of this kind of steels is irradiation embrittlement, which is manifested by a ductile-to-brittle transition temperature shift to higher temperatures after irradiation. The laboratory specimen fracture data has to be transferred to real components in order to assess the performance of these steels in the different operating and transient conditions they could find during the operation life of a fusion reactor. The specimen geometry effects and specimen size effects on measured fracture toughness need to be properly understood, taken into account and predicted with an appropriate model. The microstructure of Eurofer97 and F82H has been characterized and compared by means of optical microscopy, scanning electron microscopy, transmission electron microscopy and energy-dispersive X-ray spectroscopy in order to identify microstructural features that could play a role in the measured fracture toughness. Both steels have similar but slightly different chemical composition and final heat treatments but the prior austenitic grain size measured in F82H is approximately 8 times larger than in Eurofer97. The alloying element tantalum is added to stabilize the austenite grain size. In Eurofer97 it forms carbides of an

Vacuum Plasma Spraying W-coated Reduced Activation Structural Steels for Fusion Plasma Facing Components

Energy Technology Data Exchange (ETDEWEB)

Noh, Sanghoon; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Tungsten (W) and its alloys are considered as candidate materials for plasma facing materials of the first wall and diverter components in fusion reactor systems because of high sputtering resistance and low tritium retention in a fusion environment. Therefore, it is considered that the joining between W and reduced activation structural steels, and its evaluation, are critical issues for the development of fusion reactors. However, the joining between these materials is a very challenging process because of significant differences in their physical properties, particularly the mismatch of coefficients of thermal expansion (CTE). For instance, the CTE of pure W is known to be about 4.3Χ10{sup -6}K{sup -1}; however, that of martensitic steels reaches over three times, about 12-14Χ10{sup -6}K{sup -1} at room temperature even up to 373K. Nevertheless, several joining techniques have been developed for joining between W and structural steels, such as a vapor deposition method, brazing and diffusion bonding. Meanwhile, vacuum plasma spraying (VPS) is supposed to be one of the prospective methods to fabricate a sufficient W layer on the steel substrates because of the coating of a large area with a relatively high fabricating rate. In this study, the VPS method of W powders on reduced activation steels was employed, and its microstructure and hardness distribution were investigated. ODS ferritic steels and F82H steel were coated by VPS-W, and the microstructure and hardness distribution were investigated. A microstructure analysis revealed that pure W was successfully coated on steel substrates by the VPS process without an intermediate layer, in spite of a mismatch of the CTE between dissimilar materials. After neutron irradiation, irradiation hardening significantly occurred in the VPSW. However, the hardening of VPS-W was lesser than that of bulk W irradiated HFIR at 773K. Substrate materials, ODS ferritic steels, and F82H steel, did not show irradiation hardening

Irradiation Creep of Ferritic-Martensitic Steels EP-450, EP-823 and EI-852 Irradiated in the BN-350 Reactor over Wide Ranges of Irradiation Temperature and Dose

International Nuclear Information System (INIS)

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

2007-01-01

Full text of publication follows: Ferritic/martensitic (F/M) steels appear to be the most promising materials for advanced nuclear systems, especially for fusion reactors. Their main advantages are higher resistance to swelling and lower irradiation creep rate as has been repeatedly demonstrated in examinations of these materials after irradiation. Nevertheless, available experimental data on irradiation resistance of F/M steels are insufficient, with the greatest deficiency of data for high doses and for both low and high irradiation temperatures. From the very beginning of operation the BN-350 fast reactor has been used for irradiation of specimens of structural materials, including F/M steels. The most unique feature of BN-350 was its low inlet sodium temperature, allowing irradiation at temperatures over a very wide range of temperatures compared with the range in other fast reactors. In this paper data are presented on swelling and irradiation creep of three Russian F/M steels EP-450, EP-823 and EI-852, irradiated in experimental assemblies of the BN-350 reactor at temperatures in the range of 305-700 deg. C to doses ranging from 20 to 89 dpa. The investigation was performed using gas-pressurized creep tubes with hoop stresses in the range of 0 - 294 MPa. (authors)

The ARBOR irradiation project

International Nuclear Information System (INIS)

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

2002-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-15

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

Irradiation of steel, molybdenum and tungsten - VISA-2f; Ozracivanje celika, molibdena i volframa - VISA-2f -

Energy Technology Data Exchange (ETDEWEB)

Veljkovic, S; Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-12-15

The objective of the experiment is to study the radiation damage of steel, molybdenum and tungsten after irradiation under fast neutron flux. The sample wires of steel Mo and W will be irradiated, integral fast neutron flux should be higher than 10{sup 18} neutrons/cm{sup 2}, the temperature should be as low as possible.

The ARBOR irradiation project

Energy Technology Data Exchange (ETDEWEB)

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

2002-12-01

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

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.

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.

Thermal and mechanical behaviour of the reduced-activation-ferritic-martensitic steel EUROFER

International Nuclear Information System (INIS)

Lindau, R.; Moeslang, A.; Schirra, M.

2002-01-01

Reduced activation ferritic/martensitic (RAFM) steels are being considered for structural application in potential fusion energy systems. Based on the substantial experience with RAFM developmental steels of OPTIFER type, an industrial 3.5 tons batch of a 9CrWVTa-RAFM steel, called EUROFER 97 had been specified and ordered. A characterisation programme has been launched to determine the relevant mechanical and physical-metallurgical properties in order to qualify the steel for fusion application. The hardening, tempering and transformation behaviour of EUROFER is in good agreement with that of other RAFM-steels like OPTIFER and the Japanese industrial scale heat F82H mod. Tensile tests, performed between RT and 750 deg. C, show comparable strength and ductility values that are not strongly affected by different heat treatments and ageing at 580 and 600 deg. C up to 3300 h. Impact bending tests indicate a superior ductile to brittle transition temperature (DBTT) of EUROFER in the as-received condition compared with that of F82H mod. Creep tests between 450 and 650 deg. C up to test times of 15000 h reveal a creep strength similar to other RAFM steels like OPTIFER and F82H mod. EUROFER shows a good low-cycle fatigue behaviour with longer lifetimes than F82H mod. The deformation and softening behaviour is similar

Dynamic mechanical properties of reduced activation ferritic steels

International Nuclear Information System (INIS)

Hirose, T.; Kohyama, A.; Tanigawa, H.; Ando, M.; Jitsukawa, S.

2003-01-01

A fatigue test method by a miniaturized hourglass-shaped fatigue specimen has been developed for International Fusion Materials Irradiation Facility (IFMIF) and sufficient potential as the alternative to a conventional large specimen was presented. Furthermore, focused ion beam micro- sampling method was successfully applied to microstructural analysis on fracture process. Where, the effects of displacement damage and transmutation helium on the fatigue properties of Reduced Activation Ferritic/Martensitic Steels, RAFs, were investigated. Neutron irradiation and helium-ion-implantation at ambient temperature caused radiation hardening to degrade fatigue lifetime of F82H steel. Microstructural analysis revealed that local brittle fractures occurred at early stage of fatigue tests was the origin of the degradation.. No significant difference in fatigue life degradation was detected with and without implanted helium. This result suggests that 100 appm helium implanted has no impact on fracture life time under neutron irradiation. (author)

Mechanical properties of F82H plates with different thicknesses

Energy Technology Data Exchange (ETDEWEB)

Sakasegawa, Hideo, E-mail: sakasegawa.hideo@jaea.go.jp; Tanigawa, Hiroyasu

2016-11-01

Highlights: • Mass effect, homogeneity, and anisotropy in mechanical properties were studied. • Thickness dependence of tensile property was not observed. • Thickness dependence of Charpy impact property was observed. • Appropriate mechanical properties were obtained using an electric furnace. - Abstract: Fusion DEMO reactor requires over 11,000 tons of reduced activation ferritic/martensitic steel and it is indispensable to develop the manufacturing technology for producing large-scale components of DEMO blanket with appropriate mechanical properties. This is because mechanical properties are generally degraded with increasing production volume. In this work, we focused mechanical properties of F82H–BA12 heat which was melted in a 20 tons electric arc furnace. Plates with difference thicknesses from 18 to 100 mm{sup t} were made from its ingot through forging and hot-rolling followed by heat treatments. Tensile and Charpy impact tests were then performed on plates focusing on their homogeneity and anisotropy. From the result, their homogeneity and anisotropy were not significant. No obvious differences were observed in tensile properties between the plates with different thicknesses. However, Charpy impact property changed with increasing plate thickness, i.e. the ductile brittle transition temperature of a 100 mm{sup t} thick plate was higher than that of the other thinner plates.

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

Energy Technology Data Exchange (ETDEWEB)

Singh, B.N

2003-08-01

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 to different dose levels and at different irradiation temperatures. The resulting microstructure was characterized using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. Mechanical properties were determined by uniaxial tensile testing. Dislocation-loop interaction, formation of rafts of loops, radiation hardening and formation of 'cleared channels' were studied using different computational techniques. Experiments have shown that nano-voids are formed both in pure iron and F82H steel already at 50 deg. C. In pure iron, the formation of nano-voids is detected already at a dose level of {approx}10{sup -3} dpa. Also in iron, self-interstitial atoms were found to accumulate in the form of glissile and sessile loops; at higher dose levels, these loops led to formation of rafts of loops. Irradiation led to an increase in the yield strength, a sudden drop in the yield stress, and, at higher doses, the initiation of plastic instability immediately beyond the upper yield point. Experimental as well as the results of computer simulations are found to be consistent with the cascade induced source hardening model.

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

International Nuclear Information System (INIS)

Singh, B.N.

2003-08-01

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 to different dose levels and at different irradiation temperatures. The resulting microstructure was characterized using transmission electron microscopy, positron annihilation spectroscopy and electrical resistivity measurements. Mechanical properties were determined by uniaxial tensile testing. Dislocation-loop interaction, formation of rafts of loops, radiation hardening and formation of 'cleared channels' were studied using different computational techniques. Experiments have shown that nano-voids are formed both in pure iron and F82H steel already at 50 deg. C. In pure iron, the formation of nano-voids is detected already at a dose level of ∼10 -3 dpa. Also in iron, self-interstitial atoms were found to accumulate in the form of glissile and sessile loops; at higher dose levels, these loops led to formation of rafts of loops. Irradiation led to an increase in the yield strength, a sudden drop in the yield stress, and, at higher doses, the initiation of plastic instability immediately beyond the upper yield point. Experimental as well as the results of computer simulations are found to be consistent with the cascade induced source hardening model

Hydrogen retention in ion irradiated steels

International Nuclear Information System (INIS)

Hunn, J.D.; Lewis, M.B.; Lee, E.H.

1998-01-01

In the future 1--5 MW Spallation Neutron Source, target radiation damage will be accompanied by high levels of hydrogen and helium transmutation products. The authors have recently carried out investigations using simultaneous Fe/He,H multiple-ion implantations into 316 LN stainless steel between 50 and 350 C to simulate the type of radiation damage expected in spallation neutron sources. Hydrogen and helium were injected at appropriate energy and rate, while displacement damage was introduced by nuclear stopping of 3.5 MeV Fe + , 1 microm below the surface. Nanoindentation measurements showed a cumulative increase in hardness as a result of hydrogen and helium injection over and above the hardness increase due to the displacement damage alone. TEM investigation indicated the presence of small bubbles of the injected gases in the irradiated area. In the current experiment, the retention of hydrogen in irradiated steel was studied in order to better understand its contribution to the observed hardening. To achieve this, the deuterium isotope ( 2 H) was injected in place of natural hydrogen ( 1 H) during the implantation. Trapped deuterium was then profiled, at room temperature, using the high cross-section nuclear resonance reaction with 3 He. Results showed a surprisingly high concentration of deuterium to be retained in the irradiated steel at low temperature, especially in the presence of helium. There is indication that hydrogen retention at spallation neutron source relevant target temperatures may reach as high as 10%

Irradiation embrittlement of pressure vessel steels

International Nuclear Information System (INIS)

Brumovsky, M.; Vacek, M.

1975-01-01

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

Reversibility of U → F processes in X or UV irradiated KCl: H- systems

International Nuclear Information System (INIS)

Bressiani, Ana Helena A.

1979-01-01

Potassium Chloride crystals, pure, additively colored or hydrogenated have been exposed to X and UV radiations for correlation studies' of formation processes of defects. In hydrogenated samples, these irradiations are responsible-for a direct U→F centers conversion with the simultaneous creation of H 2 molecules besides the intrinsic formation of F/hole centers. After prolonged irradiation, F aggregate centers react with H 2 molecules forming U center pairs. As the two U centers have independent electronic transitions they add up, increasing the maximum height of the U band after F→U reversed reation. The saturation levels of defects produced by these two types of radiation are different because UV light also excites F centers while producing them. This effect is responsible by the F center aggregation and the inversion F→U decreasing the U→F efficiency. Crystals with F and U centers under F light also show F→U inversion but with higher efficiency than when under UV light. This F→U process gives back 90% of the initial U center concentration. This same F→U process was observed but at smaller rate of formation if samples containing U, F and F aggregates are left in the dark for prolonged periods of time. The same reactions take place namely H 2 molecules annihilation F aggregates. (author)

Annealing effect on restoration of irradiation steel properties

International Nuclear Information System (INIS)

Vishkarev, O.M.; Kolesova, T.N.; Myasnikova, K.P.; Pecherin, A.M.; Shamardin, V.K.

1986-01-01

The effect of temperature and annealing time on the restoration of properties of the 15Kh2NMFAA and 15Kh2MFA steels after irradiation at 285 deg with the fluence of 6x10 23 neutr/m 2 (E>0.5 MeV) is studied. Microhardness (H μ ) restoration in the irradiated 15Kh2NMFAA steel is shown to start from 350 deg C annealing temperature. The complete microhardness restoration is observed at the annealing temperature of 500 deg C for 10 hours

Effects of thermal aging on the mechanical behavior of F82H weldments

International Nuclear Information System (INIS)

Alamo, A.; Castaing, A.; Fontes, A.; Wident, P.

2000-01-01

The objective of this work is to characterize the mechanical behavior of F82H weldments, which were produced by tungsten inert gas (TIG) and electron beam (EB) processes. Tensile and impact properties were determined for both types of welds in the as-received condition and after thermal aging for 10 000 h at 400 deg. C and 550 deg. C. The mechanical properties of TIG welds and their evolution during aging is quite similar in the base metal (BM). The main difference was given by the impact energy level (upper shelf energy (USE)) of TIG joints, which is about 60% of the BM. EB welds were delivered without post-weld heat treatment. Very scattered results were obtained, particularly for impact properties

Irradiation Creep and Swelling of Russian Ferritic-Martensitic Steels Irradiated to Very High Exposures in the BN-350 Fast Reactor at 305-335 degrees C

International Nuclear Information System (INIS)

Konobeev, Yury V.; Dvoriashin, Alexander M.; Porollo, S.I.; Shulepin, S.V.; Budylkin, N.I.; Mironova, Elena G.; Garner, Francis A.

2003-01-01

Russian ferritic/martensitic (F/M) steels EP-450, EP-852 and EP-823 were irradiated in the BN-350 fast reactor in the form of gas-pressurized creep tubes. The first steel is used in Russia for hexagonal wrappers in fast reactors. The other steels were developed for compatibility with Pb-Bi coolants and serve to enhance our understanding of the general behavior of this class of steels. In an earlier paper we published data on irradiation creep of EP-450 and EP-823 at temperatures between 390 and 520C, with dpa levels ranging from 20 to 60 dpa. In the current paper new data on the irradiation creep and swelling of EP-450 and EP-852 at temperatures between 305 and 335C and doses ranging from 61 to 89 dpa are presented. Where comparisons are possible, it appears that these steels exhibit behavior that is very consistent with that of Western steels. Swelling is relatively low at high neutron exposure and confined to temperatures <420C, but may be camouflaged somewhat by precipitation-related densification. These irradiation creep studies confirm that the creep compliance of F/M steels is about one-half that of austenitic steels.

Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

Science.gov (United States)

Obraztsov, S. M.; Konobeev, Yu. V.; Birzhevoy, G. A.; Rachkov, V. I.

2006-12-01

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.

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.

Current status and recent research achievements in ferritic/martensitic steels

Science.gov (United States)

Tavassoli, A.-A. F.; Diegele, E.; Lindau, R.; Luzginova, N.; Tanigawa, H.

2014-12-01

When the austenitic stainless steel 316L(N) was selected for ITER, it was well known that it would not be suitable for DEMO and fusion reactors due to its irradiation swelling at high doses. A parallel programme to ITER collaboration already had been put in place, under an IEA fusion materials implementing agreement for the development of a low activation ferritic/martensitic steel, known for their excellent high dose irradiation swelling resistance. After extensive screening tests on different compositions of Fe-Cr alloys, the chromium range was narrowed to 7-9% and the first RAFM was industrially produced in Japan (F82H: Fe-8%Cr-2%W-TaV). All IEA partners tested this steel and contributed to its maturity. In parallel several other RAFM steels were produced in other countries. From those experiences and also for improving neutron efficiency and corrosion resistance, European Union opted for a higher chromium lower tungsten grade, Fe-9%Cr-1%W-TaV steel (Eurofer), and in 1997 ordered the first industrial heats. Other industrial heats have been produced since and characterised in different states, including irradiated up to 80 dpa. China, India, Russia, Korea and US have also produced their grades of RAFM steels, contributing to overall maturity of these steels. This paper reviews the work done on RAFM steels by the fusion materials community over the past 30 years, in particular on the Eurofer steel and its design code qualification for RCC-MRx.

Current status and recent research achievements in ferritic/martensitic steels

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A.-A.F., E-mail: farhad.tavassoli@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA/DEN/DANS/DMN, F-91191 Gif-sur-Yvette (France); Diegele, E., E-mail: eberhard.diegele@kit.edu [Karlsruhe Institut of Technology (KIT), Karlsruhe (Germany); Lindau, R., E-mail: rainer.lindau@kit.edu [Karlsruhe Institut of Technology (KIT), Karlsruhe (Germany); Luzginova, N., E-mail: Natalia.Luzginova@gmail.com [NRG-Petten, 1755 ZG Petten (Netherlands); Tanigawa, H., E-mail: tanigawa.hiroyasu@jaea.go.jp [Japan Atomic Energy Authority (JAEA), Tokai, Ibaraki, 319-1195 (Japan)

2014-12-15

When the austenitic stainless steel 316L(N) was selected for ITER, it was well known that it would not be suitable for DEMO and fusion reactors due to its irradiation swelling at high doses. A parallel programme to ITER collaboration already had been put in place, under an IEA fusion materials implementing agreement for the development of a low activation ferritic/martensitic steel, known for their excellent high dose irradiation swelling resistance. After extensive screening tests on different compositions of Fe–Cr alloys, the chromium range was narrowed to 7–9% and the first RAFM was industrially produced in Japan (F82H: Fe–8%Cr–2%W–TaV). All IEA partners tested this steel and contributed to its maturity. In parallel several other RAFM steels were produced in other countries. From those experiences and also for improving neutron efficiency and corrosion resistance, European Union opted for a higher chromium lower tungsten grade, Fe–9%Cr–1%W–TaV steel (Eurofer), and in 1997 ordered the first industrial heats. Other industrial heats have been produced since and characterised in different states, including irradiated up to 80 dpa. China, India, Russia, Korea and US have also produced their grades of RAFM steels, contributing to overall maturity of these steels. This paper reviews the work done on RAFM steels by the fusion materials community over the past 30 years, in particular on the Eurofer steel and its design code qualification for RCC-MRx.

Evaluation of irradiation hardening of proton irradiated stainless steels by nanoindentation

International Nuclear Information System (INIS)

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

2014-01-01

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

Irradiation proposition of ferritic steels in a russian reactor

International Nuclear Information System (INIS)

Seran, J.L.; Decours, J.; Levy, L.

1987-04-01

Using the low temperatures of russian reactors, a sample irradiation is proposed to study mechanical properties and swelling of martensitic steels (EM10, T91, 1.4914, HT9), ferrito-martensitic (EM12) and ferritic (F17), at temperatures lower than 400 0 C [fr

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Science.gov (United States)

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

2008-06-01

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

Irradiated accelerated corrosion of stainless steel

International Nuclear Information System (INIS)

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

2015-01-01

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

Creep behavior of double tempered 8% Cr-2% WVTa martensitic steel

International Nuclear Information System (INIS)

Tamura, Manabu; Shinozuka, Kei; Esaka, Hisao; Nowell, Matthew M.

2006-01-01

Creep testing was carried out at around 650degC for a martensitic 8Cr-2WVTa steel (F82H), which is a candidate alloy for the first wall of the fusion reactors of the Tokamak type. Rupture strength of the double tempered steel (F82HD) is lightly higher than that of simple tempered steel (F82HS). On the other hand, creep rate of F82HD is obviously smaller than that of F82HS in acceleration creep, though creep strain of F82HD in transition creep, where creep rate decreases with increasing strain, is larger than that of F82HS. Hardness of the crept H82HD decreases with increasing creep strain, which corresponded with the transmission electron microscopy (TEM) observation. On the contrary, X-ray diffraction and electron back-scattered diffraction pattern measurements show that fine sub-grains are created during transition creep. The creep curves were analyzed using an exponential type creep equation and the apparent activation energy, the activation volume and the pre-exponential factor were calculated as a function of creep strain. Then, these parameters were converted into two parameters, i.e. equivalent obstacle spacing (EOS) and mobile dislocation density parameter (MDDP). While EOS decreases with increasing creep strain, MDDP increases with increasing strain during transition creep. The decrease in EOS and the increase in either EOS or MDDP are rate-controlling factors in transition and acceleration creep, respectively. On the other hand, in case of F82HS, EOS increases and MDDP decreases during transition creep. In this case, the decrease in MDDP controls the creep rate during transition creep of F82HS. It is concluded that both EOS and MDDP are representative parameters of the change in substructure during creep. (author)

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Anon.

1991-01-01

In FY1990 the Heavy-Section Steel Irradiation (HSSI) Program was arranged into 8 tasks: (1) program management, (2) K Ic curve shift in high-copper welds, (3) K Ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K Ia curve shifts in low upper-shelf (LUS) welds, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation effects, and (8) in-service aged material evaluations. Of particular interest are the efforts in FY1990 concerning the shifts in fracture toughness and crack arrest toughness in high-copper welds, the unirradiated examination of a LUS weld from the Midland reactor, and the continued investigation into the causes of accelerated low-temperature embrittlement recently observed in RPV support steels. In the Fifth and Sixth Irradiation Series, designed to examine the shifts and possible changes in shape in the ASME K Ic and K Ia curves for two irradiated high-copper welds, it was seen that both the lower bound and mean fracture toughness shifts were greater than those of the associated Charpy-impact energies, whereas the shifts in crack arrest toughness were comparable. The irradiation-shifted fracture toughness data fell slightly below the appropriately indexed ASME K Ic curve even when it was shifted according to Revision 2 of Regulatory Guide 1.99 including its margins. The beltline weld, which was removed from the Midland reactor, fabricated by Babcock and Wilcox, Co. using Linde 80 flux, is being examined in the Tenth Irradiation Series to establish the effects of irradiation on a commercial LUS weld. A wide variation in the unirradiated fracture properties of the Midland weld were measured with values of RT NDT ranging from -22 to 54F through its thickness. In addition, a wide range of copper content from 0.21 to 0.45 wt % was found, compared to the 0.42 wt % previously reported

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Austin: austenitic steel irradiation E 145-02 Irradiation Report

International Nuclear Information System (INIS)

Genet, F.; Konrad, J.

1987-01-01

Safety measures for nuclear reactors require that the energy which might be liberated in a reactor core during an accident should be contained within the reactor pressure vessel, even after very long irradiation periods. Hence the need to know the mechanical properties at high deformation velocity of structure materials that have received irradiation damage due to their utilization. The stainless steels used in the structures of reactors undergo damage by both thermal and fast neutrons, causing important changes in the mechanical properties of these materials. Various austenitic steels available as structural materials were irradiated or are under irradiation in various reactors in order to study the evolution of the mechanical properties at high deformation velocity as a function of the irradiation damage rate. The experiment called AUSTIN (AUstenitic STeel IrradiatioN) 02 was performed by the JRC Petten Establishment on behalf of Ispra in support of the reactor safety programme

Effects of irradiation conditions and environment on the reflectivity of different steel mirrors for ITER diagnostics systems

International Nuclear Information System (INIS)

Hernandez, Teresa; Martin, Piedad; Fernandez, Pilar; Hodgson, Eric R.

2009-01-01

In this work possible degradation of the reflectivity for mirrors made from various steels subjected to ionizing radiation, at moderate temperature and in different environments (vacuum, air, or nitrogen) up to a total dose of 9 MGy, has been examined. Mirrors were prepared from conventional austenitic stainless steel (316L) and also reduced activation ferritic/martensitic (RAFM) steels (Eurofer, ODS-Eurofer, F82H), and the reflectivity studied from ultraviolet to near infrared, before and after different treatments. Under all conditions the austenitic steel mirrors only degrade slightly (<10%), however for the reduced activation steels important reflectivity degradation for wavelengths below 1000 nm are observed for the different conditions. Surface morphology and microstructure has been also investigated using scanning electron microscopy (SEM). The production of near surface nitrides is considered to be the possible cause of the optical degradation.

Embrittlement of a 17Cr ferritic steel irradiated in Phenix

International Nuclear Information System (INIS)

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

1987-01-01

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

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Corwin, W.R.

1990-08-01

The primary goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (particularly the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program includes direct continuation of irradiation studies previously conducted by the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are examined on a wide range of fracture properties. Detailed statistical analyses of the fracture data on K Ic shift of high-copper welds were performed. Analysis of the first phase of irradiated crack-arrest testing on high-copper welds was completed. Final analysis and publication of the results of the second phase of the irradiation studies on stainless steel weld-overlay cladding were completed. Determinations were made of the variations in chemistry and unirradiated RT NDT of low upper-shelf weld metal from the Midland reactor. Final analyses were performed on the Charpy impact and tensile data from the Second and Third Irradiation series on low upper-shelf welds, and the report on the series was drafted. A detailed survey of existing data on microstructural models and data bases of irradiation damage was performed, and initial development of a reaction-rate-based model was completed. 40 refs., 7 figs., 4 tabs

Neutron irradiation creep in stainless steel alloys

Energy Technology Data Exchange (ETDEWEB)

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

1994-09-01

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

Irradiation damage behavior of low alloy steel wrought and weld materials

International Nuclear Information System (INIS)

Stofanak, R.J.; Poskie, T.J.; Li, Y.Y.; Wire, G.L.

1993-01-01

A study was undertaken to evaluate the irradiation damage response of several different types of low alloy steel: vintage type ASTM A302 Grade B (A302B) plates and welds containing different Ni and Cu concentrations, 3.5% Ni steels similar to ASTM A508 Class 4, welds containing about 1% Ni (similar to type 105S), and 3.5% Ni steels with ''superclean'' composition. All materials were irradiated at several different irradiation damage levels ranging from 0.0003 to 0.06 dpa at 232C (450F). Complete Charpy V-notch impact energy transition temperature curves were generated for all materials before and after irradiation to determine transition temperature at 4IJ (30 ft-lb) or 47J (35 ft-lb) and the upper shelf energy. Irradiation damage behavior was measured by shift in Charpy 41J or 47J transition temperature (ΔTT4 41J or ΔTT 47J ) and lowering of upper shelf Charpy energy at a given irradiation damage level. It was found that chemical composition greatly influenced irradiation damage behavior; highest irradiation damage (greatest ΔTT) was found in an A302B type weld containing 1.28% Ni and 0.20% Cu while the least damage was found in 3.5% Ni, 0.05% Cu, superclean wrought materials. Combination of Ni and Cu was found to affect irradiation damage behavior at higher irradiation damage levels in the A302B welds where the 1.28% Ni, 0.20% Cu weld showed more damage than a 0.60% Ni, 0.31% Cu weld. For the 3.5% Ni steels, fabrication influenced irradiation behavior in that a silicon (Si) killed material showed greater irradiation damage than a low silicon material. In general, the 3.5% Ni materials with low copper showed less irradiation damage than the A302B materials

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.

Mechanical properties of irradiated 9Cr-2WVTa steel

International Nuclear Information System (INIS)

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

1998-01-01

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

Radiation effects on low cycle fatigue properties of reduced activation ferritic/martensitic steels

International Nuclear Information System (INIS)

Hirose, T.; Tanigawa, H.; Ando, M.; Kohyama, A.; Katoh, Y.; Narui, M.

2002-01-01

The reduced activation ferritic/martensitic steel, RAFs F82H IEA heat has been fatigue-tested at ambient temperature under diametral strain controlled conditions. In order to evaluate the effects of radiation damage and transmutation damage on fatigue characteristics, post-neutron irradiation and post-helium ion implantation fatigue tests were carried out. Fracture surfaces and fatigue crack initiation on the specimen surface were observed by SEM. Low-temperature irradiation caused an increase in stress amplitude and a reduction in fatigue lifetime corresponding to radiation hardening and loss of ductility. Neutron irradiated samples showed brittle fracture surface, and it was significant for large strain tests. On the other hand, helium implantation caused delay of cyclic softening. However, brittle crack initiation and propagation did not depend on the helium concentration profiles

46 CFR 54.25-25 - Welding of quenched and tempered steels (modifies UHT-82).

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Welding of quenched and tempered steels (modifies UHT-82... ENGINEERING PRESSURE VESSELS Construction With Carbon, Alloy, and Heat Treated Steels § 54.25-25 Welding of quenched and tempered steels (modifies UHT-82). (a) The qualification of welding procedures, welders, and...

Estimation of flow stress of radiation induced F/M steels using molecular dynamics and discrete dislocation dynamics approach

International Nuclear Information System (INIS)

More, Ameya; Dutta, B.K.; Durgaprasad, P.V.; Arya, A.K.

2012-01-01

Fe-Cr based Ferritic/Martensitic (F/M) steels are the candidate structural materials for future fusion reactors. In this work, a multi-scale approach comprising atomistic Molecular Dynamics (MD) simulations and Discrete Dislocation Dynamics (DDD) simulations are used to model the effect of irradiation dose on the flow stress of F/M steels. At the atomic scale, molecular dynamics simulations are used to study the dislocation interaction with irradiation induced defects, i.e. voids and He bubbles. Whereas, the DDD simulations are used to estimate the change in flow stress of the material as a result of irradiation hardening. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Tan, L., E-mail: tanl@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Katoh, Y. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tavassoli, A.-A.F.; Henry, J. [DMN/Dir, DEN, CEA Saclay, 91191, Gif-sur-Yvette Cedex (France); Rieth, M. [Karlsruhe Institute of Technology, Karlsruhe, 76021 (Germany); Sakasegawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Tanigawa, H. [National Institutes for Quantum and Radiological Science and Technology, Rokkasho, Aomori, 039-3212 (Japan); Huang, Q. [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2016-10-15

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 M{sub 23}C{sub 6} (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.

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

International Nuclear Information System (INIS)

Ketema, D.J.

1999-02-01

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

Irradiation response of ODS Eurofer97 steel

Energy Technology Data Exchange (ETDEWEB)

Luzginova, N.V., E-mail: luzginova@nrg.eu [Nuclear Research and Consultancy Group, Petten (Netherlands); Nolles, H.S.; Pierick, P. ten; Bakker, T.; Mutnuru, R.K.; Jong, M.; Blagoeva, D.T. [Nuclear Research and Consultancy Group, Petten (Netherlands)

2012-09-15

Oxide dispersion strengthened (ODS) Eurofer97 steel (EU batch, 0.3 wt.% of Y{sub 2}O{sub 3} particles), produced by mechanical alloying followed by hot rolling, is irradiated in the High Flux Reactor in Petten, The Netherlands at three different irradiation temperatures (300, 450 and 550 Degree-Sign C) up to nominal doses of 1 dpa and 3 dpa. The effect of neutron irradiation on the mechanical properties of ODS Eurofer97 material is investigated. It is shown that the irradiation hardening of ODS Eurofer97 steel occurs at 300 Degree-Sign C, whereas during irradiation at 450 and 550 Degree-Sign C no changes in mechanical properties are observed compared to the unirradiated material. This effect is possibly a result of the annealing of the irradiation damage at temperatures higher than 300 Degree-Sign C. The observed shifts in the Ductile to Brittle Transition Temperatures due to irradiation at different temperatures are discussed and compared with non-ODS Eurofer97 steel.

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

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

Stress corrosion cracking of highly irradiated 316 stainless steels

Energy Technology Data Exchange (ETDEWEB)

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

2001-09-01

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

Behaviour and microstructure of stainless steels irradiated in the french fast breeder reactors

International Nuclear Information System (INIS)

Dubuisson, P.; Gilbon, D.

1991-01-01

The burn-up of Fast Breeder Reactors is limited by the irradiation induced dimensional changes and mechanical properties of structural materials used for replaceable in-core components. This paper describes the behaviour improvements and also the radiation-induced microstructures of the different steels used for fuel pin cladding and wrapper tubes in French reactors. Materials of fuel pin cladding are austenitic steels whose main problem is swelling. Improvements in swelling resistance by cold-working, titanium additions and modifications of matrix (Fe-Cr-Ni) from SA 316 to CW 15-15 Ti are shown. These improvements are correlated with a best stability of microstructure under irradiation. Beneficial effects of phosphorus addition and multistabilisation (NbVTi) on radiation induced microstructure and swelling resistance are also shown. Austenitic steels used for wrapper tubes are limited both by swelling and by void embrittlement. The ferritic F17 (17Cr), ferritic-martensitic EM12 (9Cr-2MoNbV) and martensitic EM10 (9Cr-1Mo) steels present high swelling resistance. Nevertheless radiation-induced embrittlement is observed in EM12 and especially in F17. This embrittlement results from a fine and uniform radiation enhanced precipitation in ferrite grains. By contrast, the microstructure of fully martensitic EM10 steel is mush more stable and its ductile-brizzle transition temperature stays below 0 0 C. 12 figs

Study of irradiation damage structures in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

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

1997-08-01

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

Study of irradiation damage structures in austenitic stainless steels

International Nuclear Information System (INIS)

Hamada, Shozo

1997-08-01

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

Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

International Nuclear Information System (INIS)

Ghoneim, M.M.

1987-05-01

The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

Irradiation Microstructure of Austenitic Steels and Cast Steels Irradiated in the BOR-60 Reactor at 320°C

Science.gov (United States)

Yang, Yong; Chen, Yiren; Huang, Yina; Allen, Todd; Rao, Appajosula

Reactor internal components are subjected to neutron irradiation in light water reactors, and with the aging of nuclear power plants around the world, irradiation-induced material degradations are of concern for reactor internals. Irradiation-induced defects resulting from displacement damage are critical for understanding degradation in structural materials. In the present work, microstructural changes due to irradiation in austenitic stainless steels and cast steels were characterized using transmission electron microscopy. The specimens were irradiated in the BOR-60 reactor, a fast breeder reactor, up to 40 dpa at 320°C. The dose rate was approximately 9.4x10-7 dpa/s. Void swelling and irradiation defects were analyzed for these specimens. A high density of faulted loops dominated the irradiated-altered microstructures. Along with previous TEM results, a dose dependence of the defect structure was established at 320°C.

Changes in the F0F1-ATPase activity of irradiated Lactobacillus acidophilus in the presence of ceftazidime at low pH

International Nuclear Information System (INIS)

Kalantaryan, V.P.; Trchounian, A.H.; Soghomonyan, D.R.

2014-01-01

The aim of this study was the investigation of the effects of low intensity electromagnetic irradiation (EMI) at the frequencies of 51.8 and 53 GHz and of antibiotic ceftazidime on the N,N'-dicyclohexylcarbodiimide (DCCD) inhibited ATPase activity of membrane vesicles of lactic acid bacteria Lactobacillus acidophilus grown at low pH (pH 4.0 or 6.5) and assayed at the same pH. It was shown that both frequencies EMI stimulated ATPase activity of L. acidophilus grown at pH 4.0, but EMI combined with ceftazidime and DCCD decreased ATPase activity at pH 4.0 and pH 6.5. It was suggested that the F 0 F 1 -ATPase might be a target for EMI even at low pH

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).

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

International Nuclear Information System (INIS)

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

1993-07-01

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

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

International Nuclear Information System (INIS)

Kikuchi, Kenji; Kamata, Kinya; Ono, Mikinori; Kitano, Teruaki; Hayashi, Kenichi; Oigawa, Hiroyuki

2008-01-01

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 -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

Review of intense irradiation data and discussion on structural integrity

Energy Technology Data Exchange (ETDEWEB)

Suzuki, K.; Jitsukawa, S.; Okubo, N. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan)

2007-07-01

Full text of publication follows: Mechanical property data on austenitic stainless steels and F82H have been reviewed to discuss for developing structural integrity methodologies of intensely irradiated components such as first walls. The following have been already clarified: (1) Fracture ductility is still high even though tensile rupture elongation is reduced remarkably. (2) Strain-hardening occurs in true stress-logarithmic strain (true strain) relationship. Work-softening behavior observed in nominal stress-nominal strain curves is simply resulted from a reduction of work hardening rate accompanied by the increase of flow stress level by irradiation. The review lead to an innovative design concept for application to intensely irradiated components. A special consideration is given to unique feature of bending moment in developing design methodology for preventing ductile fracture of intensely irradiated materials. Another discussion is also made on how to simulate mechanical behavior of intensely irradiated components, because mechanical testing of component-wise specimens after intense irradiation is inevitable for the development of design concepts, although irradiation on such a large scale specimen seems to be almost impossible with current irradiation facilities. (authors)

Irradiation creep in ferritic steels

International Nuclear Information System (INIS)

Vandermeulen, W.; Bremaecker, A. de; Burbure, S. de; Huet, J.J.; Asbroeck, P. van

Pressurized and non-pressurized capsules of several ferritic steels have been irradiated in Rapsodie between 400 and 500 0 C up to 3.7 x 10 22 n/cm 2 (E>0.1 MeV). Results of the diameter measurements are presented and show that the total in-pile deformation is lower than for austenitic steels

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.

Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

International Nuclear Information System (INIS)

James, L.A.

1977-01-01

The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

Irradiation behavior of German PWR RPV steels under operating conditions

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Protective immunity induced in mice by F8.1 and F8.2 antigens purified from Schistosoma mansoni eggs

Directory of Open Access Journals (Sweden)

Claudia Campra Ferreira

1998-01-01

Full Text Available Schistosoma mansoni soluble egg antigens (SEA were fractionated by isoelectric focusing, resulting in 20 components, characterized by pH, absorbance and protein concentration. The higher absorbance fractions were submitted to electrophoresis, and fraction 8 (F8 presented a specific pattern of bands on its isoelectric point. Protein 3 was observed only on F8, and so, it was utilized to rabbit immunization, in order to evaluate its capacity of inducing protective immunity. IgG antibodies from rabbit anti-F8 serum were coupled to Sepharose, and used to obtain the specific antigen by affinity chromatography. This antigen, submitted to electrophoresis, presented two proteic bands (F8.1 and F8.2, which were transferred to nitrocellulose membrane (PVDF and sequenciated. The homology of F8.2 to known proteins was determined using the Basic Local Alignment Search Tool program (BLASTp. Significant homologies were obtained for the rabbit cytosolic Ca2+ uptake inhibitor, and for the bird a1-proteinase inhibitor. Immunization of mice with F8.1 and F8.2, in the presence of Corynebacterium parvum and Al(OH3 as adjuvant, induced a significant protection degree against challenge infection, as observed by the decrease on worm burden recovered from portal system.

Defects investigation in neutron irradiated reactor steels by positron annihilation

International Nuclear Information System (INIS)

Slugen, V.

2003-01-01

Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

Irradiation enhanced diffusion and irradiation creep tests in stainless steel alloys

International Nuclear Information System (INIS)

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

1977-01-01

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

Changing in tool steels wear resistance under electron irradiation

International Nuclear Information System (INIS)

Braginskaya, A.E.; Manin, V.N.; Makedonskij, A.V.; Mel'nikova, N.A.; Pakchanin, L.M.; Petrenko, P.V.

1983-01-01

The tool steels and alloys wear resistance under dry friction after electron irradiation has been studied. Electron irradiation of a wide variety of steels is shown to increase wear resistance. In this case phase composition and lattice parameters changes are observed both in matrix and carbides. The conclusion is drawn that an appreciable increase of steel wear resistance under electron irradiation can be explained both by carbide phase volume gain and changes in it's composition and the formation of carbide phase submicroscopic heterogeneities and, possibly, complexes of defects

Heavy-section steel irradiation program summary

International Nuclear Information System (INIS)

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

1992-01-01

Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict failure models of the vessel material. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations of the accompanying microstructural changes. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness (K Ic and J Ic ), crack-arrest toughness (K Ia ), ductile tearing resistance (dJ/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. (orig./GL)

Radiation-induced segregation at grain boundaries in AL-6XN stainless steels irradiated by hydrogen ions

Science.gov (United States)

Long, Yunxiang; Zheng, Zhongcheng; Guo, Liping; Zhang, Weiping; Shen, Zhenyu; Tang, Rui

2018-04-01

The effect of high concentration of hydrogen on the segregation of radiation-induced segregation (RIS) in AL-6XN stainless steels has been investigated by transmission electron microscopy (TEM) with energy-dispersive X-ray spectroscopy. Specimens were irradiated with 100 keV H2+ ions from 1 dpa to 5 dpa at 380 °C to investigated the dose dependence of grain boundary RIS. A specimen was irradiated to 5 dpa at 290 °C to study the effect of irradiation temperature. The trends of Cr depletion and Ni enrichment with irradiation dose is similar to that of other austenitic steels reported in the literatures, but the higher concentration of hydrogen made the RIS profile wider. An abnormal phenomenon that the degree of RIS increased with decreasing irradiation temperature was found, indicating that with the retention of hydrogen in the steels, temperature dependence of RIS is dominated by the quantity of retained hydrogen, rather than by thermal segregation processes.

Overview of microstructural evolution in neutron-irradiated austenitic stainless steels

International Nuclear Information System (INIS)

Maziasz, P.J.

1993-01-01

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

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......, based on the production bias model (PBM) were carried out to study the details of evolution of cavitieswith and without helium generation. The phenomena of dislocation decoration and raft formation, which are important for understanding radiation hardening and plastic flow localization, have been...... studied using the Kinetic Monte Carlo (KMC) code during arealistic dynamic irradiation of bcc iron at 300 K. Molecular dynamics (MD) simulations have been carried out to study the stress dependencies of dislocation velocity and drag coefficient for an edge dislocation decorated with small SIA loops...

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.

Neutron irradiation effects in pressure vessel steels and weldments

Energy Technology Data Exchange (ETDEWEB)

Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

1994-12-31

This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

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

Fracture toughness of irradiated and recovered vessel steels

International Nuclear Information System (INIS)

Perosanz, F.; Lapena, J.

1998-01-01

This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition. (orig.)

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Chemical coloring on stainless steel by ultrasonic irradiation.

Science.gov (United States)

Cheng, Zuohui; Xue, Yongqiang; Ju, Hongbin

2018-01-01

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

Behavior of ferritic steels irradiated by fast neutrons

International Nuclear Information System (INIS)

Erler, Jean; Maillard, Arlette; Brun, Gilbert; Lehmann, Jeanne; Dupouy, J.-M.

1979-01-01

Ferritic steels were irradiated in Rapsodie and Phenix at varying doses. The swelling and irradiation creep characteristics are reported below as are the mechanical characteristics of these materials [fr

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

Energy Technology Data Exchange (ETDEWEB)

Chang, Yongqin, E-mail: chang@ustb.edu.cn; Zhang, Jing; Li, Xiaolin; Guo, Qiang; Wan, Farong; Long, Yi

2014-12-15

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.

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.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Baard, J.H.

1996-10-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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 ∼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 ∼25% δ-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

Microstructural characterization of dissimilar welds between Incoloy 800H and 321 Austenitic Stainless Steel

Energy Technology Data Exchange (ETDEWEB)

Sayiram, G., E-mail: sayiram.g@vit.ac.in; Arivazhagan, N.

2015-04-15

In this work, the microstructural character of dissimilar welds between Incoloy 800H and 321 Stainless Steel has been discussed. The microscopic examination of the base metals, fusion zones and interfaces was characterized using an optical microscope and scanning electron microscopy. The results revealed precipitates of Ti (C, N) in the austenitic matrix along the grain boundaries of the base metals. Migration of grain boundaries in the Inconel 82 weld metal was very extensive when compared to Inconel 617 weldment. Epitaxial growth was observed in the 617 weldment which increases the strength and ductility of the weld metal. Unmixed zone near the fusion line between 321 Stainless Steel and Inconel 82 weld metal was identified. From the results, it has been concluded that Inconel 617 filler metal is a preferable choice for the joint between Incoloy 800H and 321 Stainless Steel. - Highlights: • Failure mechanisms produced by dissimilar welding of Incoloy 800H to AISI 321SS • Influence of filler wire on microstructure properties • Contemplative comparisons of metallurgical aspects of these weldments • Microstructure and chemical studies including metallography, SEM–EDS • EDS-line scan study at interface.

A study of the mechanical property changes of irradiation embrittled pressure vessel steels and their response to annealing treatments

International Nuclear Information System (INIS)

Tipping, P.; Waeber, W.B.; Mercier, O.

1991-01-01

Isochronal and isothermal heat treatments have been used to study the recovery of hardness of a neutron irradiated pressure vessel steel forging for the purposes of planning and realizing IAR (Irradiated-Annealed-Reirradiated) experiments. Charpy V notch tests have been performed to assess the toughness of the material irradiated to various fluences up to a maximum of 5 x 10 19 n/cm 2 , E>1 MeV at 290 o C with and without an intermediate annealing treatment at 450 o C x 168 h. The effect of the intermediate annealing was evident. The recovery of the upper shelf energies was strongly enhanced by a thermal ageing effect due to the annealing treatment for all fluence levels investigated compared to the irradiated condition. The transition temperature shifts exhibited a less straightforward behaviour due to the mentioned ageing effect which opposed the recovery process for this property leading to a net shift increase at lower and to a net recovery benefit at higher fluence levels. A phenomenological model description for the IAR embrittlement-recovery path is suggested. For this material and these irradiation conditions a plant life extension (PLEX) may be brought about if a specific annealing treatment is applied at a fluence level that is half the anticipated target fluence F for PLEX. In this case it was found that F>1.6 x 10 19 n/cm 2 . (author)

Reduction of upper shelf energy of highly irradiated RPV steels

Energy Technology Data Exchange (ETDEWEB)

Otaka, M.; Osaki, T. [Japan Nuclear Energy Safety Organization (Japan)

2004-07-01

It is well known that as the embrittlement due to neutron irradiation of reactor pressure vessel (RPV) steels, there is the tendency of the decrease in Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 1 x 10{sup 24} n/m{sup 2}, E>1 MeV in the OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 year plant operation for the integrity assessment of the RPVs. This paper describes the results of an atom probe tomography characterization of irradiated steels. A new form of USE prediction equation was developed based on the atom probe tomography characterization and the Charpy impact test results of the irradiated steels. And, the USE prediction equations have been determined through the regression analysis of the test reactor data combined with Japanese surveillance test data. (orig.)

Friction Stir Welding of HT9 Ferritic-Martensitic Steel: An Assessment of Microstructure and Properties

Science.gov (United States)

2013-06-01

report of FSW on a ferritic- martensitic stainless steel is the work of Chung, which applied this approach to a dissimilar weld between F82H (ferritic... martensitic ) and SUS304 (austenitic stainless ) [43]. 7 D. CORROSION OF FERRITIC/ MARTENSITIC STEELS IN HIGH TEMPERATURE MOLTEN SALT COOLANTS In...Philadelphia, PA, 1992, pp. 1267–1286, March 1990. [15] S. Rosenwasser, ―The application of martensitic stainless steels in a lifelong fusion first wall

Microstructural evolution in neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

English, C.A.; Phythian, W.J.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-01

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

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

International Nuclear Information System (INIS)

Baard, J.H.; Paardekooper, A.

1996-06-01

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

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

Science.gov (United States)

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

2017-06-01

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

Synergistic effects on dislocation loops in reduced-activation martensitic steel investigated by single and sequential hydrogen/helium ion irradiation

Energy Technology Data Exchange (ETDEWEB)

Zhang, Weiping [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Luo, Fengfeng [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Institute of Applied Physics, Jiangxi Academy of Sciences, Nanchang 330029 (China); Yu, Yanxia; Zheng, Zhongcheng; Shen, Zhenyu [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Guo, Liping, E-mail: guolp@whu.edu.cn [Hubei Nuclear Solid Physics Key Laboratory, Key Laboratory of Artificial Micro- and Nano-structures of Ministry of Education and School of Physics and Technology, Wuhan University, Wuhan 430072 (China); Ren, Yaoyao [Center for Electron Microscopy, 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)

2016-10-15

Single-beam and sequential-beam irradiations were performed to investigate the H/He synergistic effect on dislocation loops in reduced-activation ferritic/martensitic (RAFM) steels. The irradiations were carried out with 10 keV H{sup +}, 18 keV He{sup +} and 160 keV Ar{sup +}, alone and in combination at 723 K. He{sup +} single-beam irradiation induced much larger dislocation loops than that induced by both H{sup +} and Ar{sup +} single-beam irradiation. H{sup +} post-irradiation after He{sup +} irradiation further increased the size of dislocation loops, whilst He{sup +} post-irradiation or Ar{sup +} post-irradiation following H{sup +} irradiation only slightly increased the size of dislocation loops. The experiment results indicate that pre-implanted H{sup +} can drastically inhibit the growth while post-implanted H{sup +} can significantly enhance the growth of dislocation loops induced by He{sup +} irradiation. The mechanisms behind the complex synergistic phenomena between H and He and the different roles that H and He played in the growth of dislocation loops are discussed.

The irradiation performance of austenitic stainless steel clade PWR fuel rods

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-15

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

Mixed-mode I/III fracture toughness of a ferritic/martensitic stainless steel

International Nuclear Information System (INIS)

Li, Huaxin; Jones, R.H.; Gelles, D.S.; Hirth, J.P.

1993-10-01

The critical J-integrals of mode I (J IC ), mixed-mode I/III (J MC ), and mode III (J IIIC ) were examined for a ferritic stainless steel (F-82H) at ambient temperature. A determination of J MC was made using modified compact-tension specimens. Different ratios of tension/shear stress were achieved by varying the principal axis of the crack plane between 0 and 55 degrees from the load line. Results showed that J MC and tearing modulus (T M ) values varied with the crack angles and were lower than their mode I and mode III counterparts. Both the minimum J MC and T M values occurred at a crack angle between 40 and 50 degrees, where the load ratio of σ i /σ iii was 1.2 to 0.84. The J min was 240 Kj/M 2 , and ratios of J IC /J min and J IIIC /J min were 2.1 and 1.9, respectively. The morphology of fracture surfaces was consistent with the change of J MC and T M values. While the upper shelf-fracture toughness of F-82H depends on loading mode, the J min remains very high. Other important considerations include the effect of mixed-mode loading on the DBT temperature, and effects of hydrogen and irradiation on J min

Technical issues of RAFMs for the fabrication of ITER Test Blanket Module

International Nuclear Information System (INIS)

Tanigawa, Hiroyasu; Hirose, Takanori; Shiba, Kiyoyuki

2007-01-01

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

Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

International Nuclear Information System (INIS)

McGowan, J.J.; Nanstad, R.K.; Thoms, K.R.; Menke, B.H.

1985-01-01

This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs

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

International Nuclear Information System (INIS)

Schubert, L.E.; Hamilton, M.L.; Gelles, D.S.

1997-01-01

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 ∼67 dpa and at 370 degrees C to ∼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 ∼67 dpa than after irradiation at 370 degrees C to ∼15 dpa

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.

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.

Heavy-Section Steel Irradiation Program

Energy Technology Data Exchange (ETDEWEB)

Rosseel, T.M.

2000-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established.

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Rosseel, T.M.

2000-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established

Characterization of ion beam irradiated 304 stainless steel utilizing nanoindentation and Laue microdiffraction

Energy Technology Data Exchange (ETDEWEB)

Lupinacci, A. [Department of Materials Science and Engineering, University of California, Berkeley, CA (United States); Chen, K., E-mail: kchenlbl@gmail.com [Center for Advancing Materials Performance from the Nanoscale (CAMP-Nano), State Key Laboratory for Mechanical Behavior of Materials, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Li, Y. [Center for Advancing Materials Performance from the Nanoscale (CAMP-Nano), State Key Laboratory for Mechanical Behavior of Materials, Xi’an Jiaotong University, Xi’an, Shaanxi 710049 (China); Kunz, M. [Advanced Light Source, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Jiao, Z.; Was, G.S. [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States); Abad, M.D. [Department of Nuclear Engineering, University of California, Berkeley, CA (United States); Minor, A.M. [Department of Materials Science and Engineering, University of California, Berkeley, CA (United States); National Center for Electron Microscopy, The Molecular Foundry, Lawrence Berkeley National Laboratory, Berkeley, CA (United States); Hosemann, P., E-mail: Peterh@berkeley.edu [Department of Nuclear Engineering, University of California, Berkeley, CA (United States)

2015-03-15

Characterizing irradiation damage in materials utilized in light water reactors is critical for both material development and application reliability. Here we use both nanoindentation and Laue microdiffraction to characterize both the mechanical response and microstructure evolution due to irradiation. Two different irradiation conditions were considered in 304 stainless steel: 1 dpa and 10 dpa. In addition, an annealed condition of the 10 dpa specimen for 1 h at 500 °C was evaluated. Nanoindentation revealed an increase in hardness due to irradiation and also revealed that hardness saturated in the 10 dpa case. Broadening using Laue microdiffraction peaks indicates a significant plastic deformation in the irradiated area that is in good agreement with both the SRIM calculations and the nanoindentation results.

Development of PIE techniques for irradiated LWR pressure vessel steels

International Nuclear Information System (INIS)

Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

1999-01-01

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)

Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation

International Nuclear Information System (INIS)

Klueh, R.L.; Vitek, J.M.; Corwin, W.R.; Alexander, D.J.

1987-01-01

Miniature Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 1 and 2% Ni were irradiated in the High-Flux Isotope Reactor (HFIR) at 50 0 C to displacement damage levels of up to 9 dpa. Nickel was added to study the effect of transmutation helium. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT). The 9Cr-1MoVNb steels, with and without nickel, showed a larger shift than the 12Cr-1MoVW steels, with and without nickel. The results indicated that helium also increased the DBTT. The same steels were previously irradiated at higher temperatures. From the present and past tests, the effect of irradiation temperature on the DBTT behavior can be evaluated. For the 9Cr-1MoVNb steel, there is a continuous decrease in the magnitude of the DBTT increase up to an irradiation temperature of about 400 0 C, after which the shift drops rapidly to zero at about 450 0 C. The DBTT of the 12Cr-1MoVW steel shows a maximum increase at an irradiation temperature of about 400 0 C and less of an increase at either higher or lower irradiation temperatures

Influence of tempering on mechanical properties of ferritic martensitic steels

International Nuclear Information System (INIS)

Chun, Y. B.; Han, C. H.; Choi, B. K.; Lee, D. W.; Kim, T. K.; Jeong, Y. H.; Cho, S.

2012-01-01

In the mid-1980s research programs for development of low activation materials began. This is based on the US Nuclear Regulatory Commission Guidelines (10CFR part 61) that were developed to reduce long-lived radioactive isotopes, which allows nuclear reactor waste to be disposed of by shallow land burial when removed from service. Development of low activation materials is also key issue in nuclear fusion systems, as the structural components can became radioactive due to nuclear transmutation caused by exposure to high dose neutron irradiation. Reduced-activation ferritic martensitic (RAFM) steels have been developed in the leading countries in nuclear fusion technology, and are now being considered as primary candidate material for the test blanket module (TBM) in the international thermonuclear experiment reactor (ITER). RAFM steels developed so far (e.g., EUROFER 97 and F82H) meet the requirement for structural application in the ITER. However, if such alloys are used in the DEMO or commercial fusion reactor is still unclear, as the reactors are designed to operate under much severe conditions (i.e., higher outlet coolant temperature and neutron fluences). Such harsh operating conditions lead to development of RAFM steels with better creep and irradiation resistances. Mechanical properties of RAFM steels are strongly affected by microstructural features including the distribution, size and type of precipitates, dislocation density and grain size. For a given composition, such microstructural characteristics are determined mainly by thermo-mechanical process employed to fabricate the final product, and accordingly a final heat treatment, i.e., tempering is the key step to control the microstructure and mechanical properties. In the present work, we investigated mechanical properties of the RAFM steels with a particular attention being paid to effects of tempering on impact and creep properties

Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

International Nuclear Information System (INIS)

Lott, R.G.; Freyer, P.D.

1996-01-01

The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior

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.

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

International Nuclear Information System (INIS)

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

1996-06-01

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

Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

Microstructural evolution in austenitic stainless steel irradiated with triple-beam

Energy Technology Data Exchange (ETDEWEB)

Hamada, Shozo; Miwa, Yukio; Yamaki, Daiju [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yichuan, Zhang

1997-03-01

An austenitic stainless steel was simultaneously irradiated with nickel, helium and hydrogen ions at the temperature range of 573-673 K. The damage level and injected concentration of He and H ions in the triple-beam irradiated region are 57 dpa, 19000 and 18000 at.ppm, respectively. Following to irradiation, the cross sectional observation normal to the incident surface of the specimen was carried out with a transmission electron microscope. Two bands parallel to the incident surface were observed in the irradiated specimen, which consist of dislocation loops and lines of high number density. These locate in the range of the depth of 0.4 to 1.3 {mu}m and 1.8 to 2.4 {mu}m from the incident surface, respectively. The region between two bands, which corresponds to the triple beam irradiated region, shows very low number density of dislocations than that in each band. Observation with higher magnification of this region shows that fine cavities with high number density uniformly distribute in the matrix. (author)

Study of irradiation effects in austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

Kinetics of annealing of irradiated surveillance pressure vessel steel

International Nuclear Information System (INIS)

Harvey, D.J.; Wechsler, M.S.

1982-01-01

Indentation hardness measurements as a function of annealing were made on broken halves of Charpy impact surveillance samples. The samples had been irradiated in commercial power reactors to a neutron fluence of approximately 1 x 10 18 neutrons per cm 2 , E > 1 MeV, at a temperature of about 300 0 C (570 0 F). Results are reported for the weld metal, which showed greater radiation hardening than the base plate or heat-affected zone material. Isochronal and isothermal anneals were conducted on the irradiated surveillance samples and on unirradiated control samples. No hardness changes upon annealing occurred for the control samples. The recovery in hardness for the irradiated samples took place mostly between 400 and 500 0 C. Based on the Meechan-Brinkman method of analysis, the activation energy for annealing was found to be 0.60 +- 0.06 eV. According to computer simulation calculations of Beeler, the activation energy for migration of vacancies in alpha iron is about 0.67 eV. Therefore, the results of this preliminary study appear to be consistent with a mechanism of annealing of radiation damage in pressure vessel steels based on the migration of radiation-produced lattice vacancies

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

International Nuclear Information System (INIS)

Dalton, J.H.

1978-01-01

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

Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels

International Nuclear Information System (INIS)

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

1997-01-01

The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K Jc , predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material)

Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

International Nuclear Information System (INIS)

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

1992-01-01

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

Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

Energy Technology Data Exchange (ETDEWEB)

Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

2015-12-15

Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

Creep strength of reduced activation ferritic/martensitic steel Eurofer'97

International Nuclear Information System (INIS)

Fernandez, P.; Lancha, A.M.; Lapena, J.; Lindau, R.; Rieth, M.; Schirra, M.

2005-01-01

Creep rupture strength of tempered martensitic steel Eurofer'97 has been investigated. Different products form (plate and bar) have been tested in the temperature range from 450 deg. C to 650 deg. C at different loads. No significant differences in the creep rupture properties have been found between the studied product forms. The Eurofer'97 has shown adequate creep rupture strength levels at short creep rupture tests, similar to those of the F-82 H mod. steel. However, for long testing times (>9000 h) the results available up to now at 500 deg. C and 550 deg. C seem to indicate a change in the creep degradation mechanism

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-09-01

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

Metallurgical characterization of the reduced activation ferritic/martensitic steel Eurofer'97 on as-received condition

International Nuclear Information System (INIS)

Fernandez, P.; Lancha, A.M.; Lapena, J.; Hernandez-Mayoral, M.

2001-01-01

A new European reduced activation ferrous alloy (denominated Eurofer'97) developed as possible first wall and breeder blanket structural material for fusion applications is being characterized. In this paper, activities specially focussed to investigate the microstructural and mechanical properties of this material on the as-received state (normalized at 980 degree sign C/27' plus tempered at 760 degree sign C/90'/air cooled) are presented. Chemical analyses, a detailed microstructural study, hardness, tensile and Charpy tests have been carried out and are compared to the reduced activation material F-82H modified previously studied. The results show that the Eurofer'97 is a fully martensitic steel free of δ-ferrite with similar tensile and better impact properties than the F-82H modified steel. Two types of carbides have been observed in the Eurofer'97, namely, Cr rich precipitates and Ta/V rich precipitates, tentatively identified as M 23 C 6 type and (Ta,V)C type, respectively

Pressure vessel steels: influence of chemical composition on irradiation sensitivity

International Nuclear Information System (INIS)

Ghoniem, M.M.; Hammad, F.H.

1998-01-01

Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Response of neutron-irradiated RPV steels to thermal annealing

International Nuclear Information System (INIS)

Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

1997-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels

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

Energy Technology Data Exchange (ETDEWEB)

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

1967-09-15

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

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

International Nuclear Information System (INIS)

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

1967-09-01

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

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.

Martensitic transformation induced by irradiation and deformation in stainless steels

International Nuclear Information System (INIS)

Maksimkin, O.P.

1997-01-01

In the present work the peculiarities of martensite γ → α , (γ → ε → α , ) transformation in the steels with a low stacking fault energy (12Cr18Ni10T, Cr15AG14) irradiated by neutrons, α-particles and electrons (pulse and stationary) and then deformed with the various strain rates in the temperature range - 20 - 1000 C are considered. It is established by the electron-microscope research that the phase γ → α ' transition in irradiated and deformed steels is observed on the definite stage of evolution of the dislocation structure (after the cell formation) and the martensite formation preferentially occurs on a stacking fault aggregation. The regularities of the irradiation by high energy particles effect on the formation parameters and martensite α , -phase accumulation kinetics ones and also their role in forming of the strength and ductile properties in steels are analysed. (A.A.D.)

Brittle and ductile rupture of 16MND5 steel. Irradiation effect

International Nuclear Information System (INIS)

Al Mundheri, M.; Soulat, P.; Pineau, A.

1986-06-01

Toughness tests have been made on 16MND5 steel (A508Cl3 steel) - before and after irradiation at 290 0 C (3.10 19 n/cm 2 , E > 1 MeV). It is shown that toughness is lowered following the irradiation and that it is a decreasing function of the thickness of the test pieces. In parallel, tests on three geometries of entailed specimens, prepared in the non-irradiated material, have been made at different temperatures to apply the methodology of local approach of ductile-brittle rupture [fr

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Formation of ultra-fine grained SUS316L steels by ball-milling and their mechanical properties after neutron irradiation

International Nuclear Information System (INIS)

Zheng, Y.J.; Yamasaki, T.; Fukami, T.; Terasawa, M.; Mitamura, T.

2003-01-01

In order to overcome the irradiation embrittlement in austenitic stainless steels, ultra-fine grained SUS316L steels with very fine TiC particles have been developed. The SUS316L-TiC nanocomposite powders having 1.0 to 2.0 mass% TiC were prepared by ball-milling SUS316L-TiC powder mixtures for 125 h in an argon gas atmosphere. The milled powders were consolidated by hot isostatic pressing (HIP) under a pressure of 200 MPa at temperatures between 700 and 1000 C, and the bulk materials with grain sizes between 100 and 400 nm have been produced. The possibility of using fine-grained TiC particles to pin grain boundaries and thereby maintain the ultra-fine grained structures has been discussed. In order to clarify the effects of the neutron irradiation on mechanical properties of the ultra-fine grained SUS316L steels, Vickers microhardness measurements were performed before and after the irradiation of 1.14 x 10 23 n/m 2 and 1.14 x 10 24 n/m 2 . The hardness increased with increasing the dose of the irradiation. However, these increasing rates of the ultra-fine grained steels were much smaller than those of the coarse-grained SUS316L steels having grain sizes between 13 and 50 μm. (orig.)

Hardness of AISI type 410 martensitic steels after high temperature irradiation via nanoindentation

Science.gov (United States)

Waseem, Owais Ahmed; Jeong, Jong-Ryul; Park, Byong-Guk; Maeng, Cheol-Soo; Lee, Myoung-Goo; Ryu, Ho Jin

2017-11-01

The hardness of irradiated AISI type 410 martensitic steel, which is utilized in structural and magnetic components of nuclear power plants, is investigated in this study. Proton irradiation of AISI type 410 martensitic steel samples was carried out by exposing the samples to 3 MeV protons up to a 1.0 × 1017 p/cm2 fluence level at a representative nuclear reactor coolant temperature of 350 °C. The assessment of deleterious effects of irradiation on the micro-structure and mechanical behavior of the AISI type 410 martensitic steel samples via transmission electron microscopy-energy dispersive spectroscopy and cross-sectional nano-indentation showed no significant variation in the microscopic or mechanical characteristics. These results ensure the integrity of the structural and magnetic components of nuclear reactors made of AISI type 410 martensitic steel under high-temperature irradiation damage levels up to approximately 5.2 × 10-3 dpa.

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.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

On the way to high resolution TEM characterization of dual ion beam irradiated ODS steels

International Nuclear Information System (INIS)

Hsiung, L.; Tumey, S.; Fluss, M. J.; King, W.; Marian, J.; Kuntz, J.; Dasher, B. El; Serruys, Y.; Willaime, F.; Kimura, A.

2009-01-01

Fission and fusion energy application of ODS steels while appearing promising requires that many key science issues be resolved. Among these issues are our incomplete understanding of the effect of irradiation on low-temperature fracture properties, the role of fusion relevant helium and hydrogen transmutation gases on the deformation and fracture of irradiated material at low and high temperatures, radiation-induced solute segregation and phase stability, mechanisms of swelling suppression in ODS steels, and the effects of radiation damage on localized deformation. While planning to focus on all these issues we are particularly interested in the atomic scale mechanism by which helium is mitigated by the nano scale particles. In order to obtain insight we are performing analytical transmission electron microscopy (AEM), high resolution electron microscopy (HRTEM) to investigate micro-structural and micro-compositional changes and property alterations of Fe-Cr ferritic/martensitic and ODS steels driven by temperature and ion-beam irradiation with Fe, H, and He. As a beginning to a collaboration between LLNL and CEA-Saclay, we have carried out an irradiation of four specimens, Fe, Fe14%Cr, and two ODS steels (14% Cr and 16% Cr) using the dual beam facility at CEA-Saclay (JANNuS). An Fe 8+ beam was implanted at 24 MeV and helium was implanted through a degrader wheel with energies between 1.7 MeV and 1.3 MeV. The nominal radiation parameters were 40 to 25 DPA, 10 to 25 appm He/DPA ratio, and specimen temperatures of ∼425 deg. C. Our goal is to compare the evolved microstructure with respect to the accumulation of helium at or near the particle matrix interface. Preparatory to this first study we have made many hi-resolution analyses of the nano-particles in the two ODS steels which serve as a base line for comparison with the TEM post irradiation examination reported here. These base line studies are reported separately at this conference. (author)

Material property changes of stainless steels under PWR irradiation

International Nuclear Information System (INIS)

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

2009-01-01

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

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

International Nuclear Information System (INIS)

Milasin, N.

1964-05-01

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, σ y =σ i +k y d -1/2 between lower yield stress, σ y , and grain size, 2d, the information about the effect of irradiation on the parameters σ i and k y is obtained. Taking as a base interpretation of σ i and k y given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of the experimental results obtained the relative microstructure and

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.

Theoretical description of the influence of neutron irradiation on viscoplastic properties of mild steel

International Nuclear Information System (INIS)

Pecherski, R.

1978-01-01

The physical bases of plastic deformation of mild steel are described. The influence of neutron irradiation on the change of mechanisms of plastic deformation is discussed in detail. Constitutive equations of viscoplasticity for irradiated mild steel are given. The problem of thickwalled viscoplastic spherical tank irradiated by neutrons is studied. (Z.R.)

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

International Nuclear Information System (INIS)

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

1984-01-01

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

Precipitation response of austenitic stainless steel to simulated fusion irradiation

International Nuclear Information System (INIS)

Maziasz, P.J.

1978-01-01

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

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

International Nuclear Information System (INIS)

Gupta, Jyoti

2016-01-01

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

Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging

International Nuclear Information System (INIS)

Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

1988-01-01

Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1x10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6 C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6 C in JPCA-2. (orig.)

Conductivity of γ-irradiated LiF

International Nuclear Information System (INIS)

Guckelsberger, K.

1974-01-01

The conductivity of pure monocrystalline LiF, γ-irradiated, was measured in the temperature range 0.3-100 deg K. The F centers produced by irradiation disperse phonons like point defects, the corresponding fluorine atoms on inter-lattice sites appear as clusters of about 10 A diameter, and high irradiation doses (above 10 8 R) give rise to extensive perturbation zones. Each of these 3 defects influences heat conduction in a characteristic temperature range and may thus be analyzed during irradiation and consecutive recovery [fr

Irradiation-induced creep in 316 and 304L stainless steels

International Nuclear Information System (INIS)

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

1977-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-01

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

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

International Nuclear Information System (INIS)

Filacchioni, G.; Montanari, R.; Tata, M.E.; Pilloni, L.

2002-01-01

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

Hardening and microstructural evolution of A533b steels irradiated with Fe ions and electrons

Energy Technology Data Exchange (ETDEWEB)

Watanabe, H., E-mail: watanabe@riam.kyushu-u.ac.jp [Research Institute for Applied Mechanics, Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Arase, S. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Yamamoto, T.; Wells, P. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States); Onishi, T. [Interdisciplinary Graduate School of Kyushu University, 6-1, Kasuga-kouenn, Kasugashi, Fukuoka, 816-8580 (Japan); Odette, G.R. [Dept. Chemical Engineering, UCSB Engineering II, RM3357, Santa Barbara, CA, 93106-5080 (United States)

2016-04-01

Radiation hardening and embrittlement of A533B steels is heavily dependent on the Cu content. In this study, to investigate the effect of copper on the microstructural evolution of these materials, A533B steels with different Cu levels were irradiated with 2.4 MeV Fe ions and 1.0 MeV electrons. Ion irradiation was performed from room temperature (RT) to 350 °C with doses up to 1 dpa. At RT and 290 °C, low dose (<0.1 dpa) hardening trend corresponded with ΔH ∝ (dpa){sup n}, with n initially approximately 0.5 and consistent with a barrier hardening mechanism, but saturating at ≈0.1 dpa. At higher dose levels, the radiation-induced hardening exhibited a strong Cu content dependence at 290 °C, but not at 350 °C. Electron irradiation using high-voltage electron microscopy revealed the growth of interstitial-type dislocation loops and enrichment of Ni, Mn, and Si in the vicinities of pre-existing dislocations at doses for which the radiation-induced hardness due to ion irradiation was prominent.

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

International Nuclear Information System (INIS)

Nystrand, A.C.

1994-02-01

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

Microstructure and grain size effects on irradiation hardening of low carbon steel for reactor tanks

Energy Technology Data Exchange (ETDEWEB)

Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-05-15

Irradiation hardening of steel for reactor pressure vessels has been studied extensively during the past few years. A great number of experimental results concerning the behaviour of these steels in the radiation field and several review papers (1,2) have been published. Most of the papers deal with the effects of specific metallurgical factors or irradiation conditions (temperature, flux) on irradiation hardening and embrittlement. In addition, a number of experiments are performed to give evidence on the mechanism of irradiation hardening of these steels. However, this mechanism is still unknown due to the complexity of steel as a system. Among different methods used in radiation damage studies, the changes of mechanical properties have been mainly investigated. By using Hall-Petch's empirical relation, {sigma}{sub y}={sigma}{sub i}+k{sub y} d{sup -1/2} between lower yield stress, {sigma}{sub y}, and grain size, 2d, the information about the effect of irradiation on the parameters {sigma}{sub i} and k{sub y} is obtained. Taking as a base interpretation of {sigma}{sub i} and k{sub y} given by Petch and his co-workers it has been concluded that radiation does not change the stress to start slip but that it increase the friction that opposes the passage of free dislocations across a slip plane. In attempting to apply Hall-Petch's relation to one unirradiated ferritic steel with a carbon content higher than 0.15% some difficulties were encountered. The results obtained indicate that the influence of grain size can not be isolated from other factors introduced by the treatments used to produce different grain sizes. This paper deals with a similar problem in the case of irradiated steel. The results obtained give the changes of the mechanical properties of steel in neutron irradiation field as a function of microstructure and grain size. In addition, the mechanical properties of irradiated steel are measured after annealing at 150 deg C and 450 deg C. On the basis of

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.

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

International Nuclear Information System (INIS)

Klaar, H.J.

1975-01-01

As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

Irradiation effects of 11 MeV protons on ferritic steels

International Nuclear Information System (INIS)

Hamaguchi, Yoshikazu; Kuwano, Hisashi; Misawa, Toshihei

1985-01-01

It is considered that ferritic/martensitic steels are the candidate of the first wall materials for future fusion reactors. The most serious problem in the candidate materials is the loss of ductility due to the elevation of ductile-brittle transition temperature by the high dpa irradiation of neutrons. 14 MeV neutrons produced by D-T reaction cause high dpa damage and also produce large quantity of helium and hydrogen atoms in first wall materials. Those gas atoms also play an important role in the embrittlement of steels. The main purpose of this work was to simulate the behavior of hydrogen produced by the transmutation in the mechanical properties of ferritic steels when they were irradiated with 11 MeV protons. The experimental procedure and the results of hardness, the broadening of x-ray diffraction lines, Moessbauer spectroscopy and small punch test are reported. High energy protons of 10 - 20 MeV are suitable to the simulation experiment of 14 MeV neutron radiation damage. But the production of the active nuclei emitting high energy gamma ray and having long life, Co-56, is the most serious problem. Another difficulty is the control of irradiation temperature. A small irradiation chamber must be developed. (Kako, I.)

Study of 316 stainless steel swelling due to neutron irradiation

International Nuclear Information System (INIS)

Furutani, Gen; Konishi, Takao

2000-01-01

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

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

International Nuclear Information System (INIS)

Volgin, Alexandre

2012-01-01

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

Friction stir welding of F/M ODS steel plug and F/M steel tube

Energy Technology Data Exchange (ETDEWEB)

Kang, Suk Hoon, E-mail: shkang77@kaeri.re.kr [Nuclear Materials Division, Korea Atomic Energy Research Institute (Korea, Republic of); Vasudevan, M. [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Noh, Sanghoon; Jin, Hyun Ju; Jang, Jinsung; Kim, Tae Kyu [Nuclear Materials Division, Korea Atomic Energy Research Institute (Korea, Republic of)

2016-11-01

Highlights: • Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. • The curvature and smaller thickness of tube was the major limitation for applying FSW method, it was solved using specially designed jig. • Considerable hardening occurs in the joint because the cooling rate was sufficient to reproduce a martensitic microstructure. • The measured hoop strength of the FSWed joint was 70–90 MPa, the value was at around 70% of the tube. - Abstract: Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. The dimensions of the tube included outer diameter of 7 mm, wall thickness of 0.5 mm. The objective was to find suitable process variables for gaining enough frictional heat from those thin and curved pieces. A specially designed jig was used for stabilization and slow rotation of tube during FSW. Additionally, the plug was designed to overlap the tube. Inconel 718 was used as FSW tool, the diameter was 3.5 mm. The adequate rotation speed of the tool and jig were 1200 rpm and 1.5 rpm, respectively. The joining was successfully accomplished using above combination, showing a good possibility. The hoop stress tests of joint were conducted by blowing Ar gas into the tube, the flow rate of gas was 10 MPa/min. The measured hoop stress was 70–90 MPa, the value was at around 70% of the tube.

Friction stir welding of F/M ODS steel plug and F/M steel tube

International Nuclear Information System (INIS)

Kang, Suk Hoon; Vasudevan, M.; Noh, Sanghoon; Jin, Hyun Ju; Jang, Jinsung; Kim, Tae Kyu

2016-01-01

Highlights: • Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. • The curvature and smaller thickness of tube was the major limitation for applying FSW method, it was solved using specially designed jig. • Considerable hardening occurs in the joint because the cooling rate was sufficient to reproduce a martensitic microstructure. • The measured hoop strength of the FSWed joint was 70–90 MPa, the value was at around 70% of the tube. - Abstract: Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. The dimensions of the tube included outer diameter of 7 mm, wall thickness of 0.5 mm. The objective was to find suitable process variables for gaining enough frictional heat from those thin and curved pieces. A specially designed jig was used for stabilization and slow rotation of tube during FSW. Additionally, the plug was designed to overlap the tube. Inconel 718 was used as FSW tool, the diameter was 3.5 mm. The adequate rotation speed of the tool and jig were 1200 rpm and 1.5 rpm, respectively. The joining was successfully accomplished using above combination, showing a good possibility. The hoop stress tests of joint were conducted by blowing Ar gas into the tube, the flow rate of gas was 10 MPa/min. The measured hoop stress was 70–90 MPa, the value was at around 70% of the tube.

Detection of myocardial metabolic abnormalities by 18F-FDG PET/CT and corresponding pathological changes in beagles with local heart irradiation

Energy Technology Data Exchange (ETDEWEB)

Yan, Rul [Nursing College of Shanxi Medical University, Taiyuan (China); Song, Jianbo; Wu, Zhi Fang; Liu, Jian Zhang; Hao, Xin Zhong; Li, Sijin [Dept. of Nuclear Medicine, First Hospital of Shanxi Medical University, Taiyuan (China); Guo, Min [Dept. of Cardiology, First Hospital of Shanxi Medical University, Taiyuan (China); Li, Jianguo [Dept. of Radiological and Environmental Medicine, China Institute for Radiation Protection, Taiyuan (China)

2015-08-15

To determine the efficacy of 18F-fluorodeoxyglucose positron emission tomography/computed tomography (18F-FDG PET/CT) in the detection of radiation-induced myocardial damage in beagles by comparing two pre-scan preparation protocols as well as to determine the correlation between abnormal myocardial FDG uptake and pathological findings. The anterior myocardium of 12 beagles received radiotherapy locally with a single X-ray dose of 20 Gy. 18F-FDG cardiac PET/CT was performed at baseline and 3 months after radiation. Twelve beagles underwent two protocols before PET/CT: 12 hours of fasting (12H-F), 12H-F followed by a high-fat diet (F-HFD). Regions of interest were drawn on the irradiation and the non-irradiation fields to obtain their maximal standardized uptake values (SUVmax). Then the ratio of the SUV of the irradiation to the non-irradiation fields (INR) was computed. Histopathological changes were identified by light and electron microscopy. Using the 12H-F protocol, the average INRs were 1.18 ± 0.10 and 1.41 ± 0.18 before and after irradiation, respectively (p = 0.021). Using the F-HFD protocol, the average INRs were 0.99 ± 0.15 and 2.54 ± 0.43, respectively (p < 0.001). High FDG uptake in irradiation field was detected in 33.3% (4/12) of 12H-F protocol and 83.3% (10/12) of F-HFD protocol in visual analysis, respectively (p = 0.031). The pathology of the irradiated myocardium showed obvious perivascular fibrosis and changes in mitochondrial vacuoles. High FDG uptake in an irradiated field may be related with radiation-induced myocardial damage resulting from microvascular damage and mitochondrial injury. An F-HFD preparation protocol used before obtaining PET/CT can improve the sensitivity of the detection of cardiotoxicity associated with radiotherapy.

Carbon Contamination During Ion Irradiation - Accurate Detection and Characterization of its Effect on Microstructure of Ferritic/Martensitic Steels

Energy Technology Data Exchange (ETDEWEB)

Wang, Jing; Toloczko, Mychailo B.; Kruska, Karen; Schreiber, Daniel K.; Edwards, Danny J.; Zhu, Zihua; Zhang, Jiandong

2017-11-17

Accelerator-based ion beam techniques have been used to study radiation effects in materials for decades. Although carbon contamination induced by ion beam in target materials is a well-known issue, it has not been fully characterized nor quantified for studies in ferritic/martensitic (F/M) steels that are candidate materials for applications such as core structural components in advanced nuclear reactors. It is an especially important issue for this class of material because of the effect of carbon level on precipitate formation. In this paper, the ability to quantify carbon contamination using three common techniques, namely time-of-flight secondary ion mass spectroscopy (ToF-SIMS), atom probe tomography (APT) and transmission electron microscopy (TEM) is compared. Their effectiveness and short-comings in determining carbon contamination will be presented and discussed. The corresponding microstructural changes related to carbon contamination in ion irradiated F/M steels are also presented and briefly discussed.

Heavy-Section Steel Irradiation Program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. It is imperative to understand and predict the capabilities and limitations of its integrity. It is particularly vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. The Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Results from HSSI studies provide information needed to aid in resolving major regulatory issues facing the USNRC which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf (LUS) welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. Embrittlement modeling studies have shown that the time or dose required for the point defect concentrations, which ultimately contribute to irradiation embrittlement, to reach their steady state values can be comparable to the component lifetime or to the duration of an irradiation experiment

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

International Nuclear Information System (INIS)

Corwin, W.R.

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

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

R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

2000-01-01

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 thye 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)

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.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Characterization of matrix damage in ion-irradiated reactor vessel steel

International Nuclear Information System (INIS)

Fujii, Katsuhiko; Fukuya, Koji

2004-01-01

Exact nature of the matrix damage, that is one of radiation-induced nano-scale microstructural features causing radiation embrittlement of reactor vessel, in irradiated commercial steels has not been clarified yet by direct characterization using transmission electron microscopy (TEM). We designed a new preparation method of TEM observation samples and applied it to the direct TEM observation of the matrix damage in the commercial steel samples irradiated by ions. The simulation irradiation was carried out by 3 MeV Ni 2+ ion to a dose of 1 dpa at 290degC. Thin foil specimens for TEM observation were prepared using the modified focused ion beam method. A weak-beam TEM study was carried out for the observation of matrix damage in the samples. Results of this first detailed observation of the matrix damage in the irradiated commercial steel show that it is consisted of small dislocation loops. The observed and analyzed dislocation loops have Burgers vectors b = a , and a mean image size and the number density are 2.5 nm and about 1 x 10 22 m -3 , respectively. In this experiment, all of the observed dislocation loops were too small to determine the vacancy or interstitial nature of the dislocation loops directly. Although it is an indirect method, post-irradiation annealing was used to infer the loop nature. Most of dislocation loops were stable after the annealing at 400degC for 30 min. This result suggests that their nature is interstitial. (author)

Mechanical Performance of Ferritic Martensitic Steels for High Dose Applications in Advanced Nuclear Reactors

Science.gov (United States)

Anderoglu, Osman; Byun, Thak Sang; Toloczko, Mychailo; Maloy, Stuart A.

2013-01-01

Ferritic/martensitic (F/M) steels are considered for core applications and pressure vessels in Generation IV reactors as well as first walls and blankets for fusion reactors. There are significant scientific data on testing and industrial experience in making this class of alloys worldwide. This experience makes F/M steels an attractive candidate. In this article, tensile behavior, fracture toughness and impact property, and creep behavior of the F/M steels under neutron irradiations to high doses with a focus on high Cr content (8 to 12) are reviewed. Tensile properties are very sensitive to irradiation temperature. Increase in yield and tensile strength (hardening) is accompanied with a loss of ductility and starts at very low doses under irradiation. The degradation of mechanical properties is most pronounced at martensitic steels exhibit a high fracture toughness after irradiation at all temperatures even below 673 K (400 °C), except when tested at room temperature after irradiations below 673 K (400 °C), which shows a significant reduction in fracture toughness. Creep studies showed that for the range of expected stresses in a reactor environment, the stress exponent is expected to be approximately one and the steady state creep rate in the absence of swelling is usually better than austenitic stainless steels both in terms of the creep rate and the temperature sensitivity of creep. In short, F/M steels show excellent promise for high dose applications in nuclear reactors.

Fracture toughness of irradiated stainless steel alloys

International Nuclear Information System (INIS)

Mills, W.J.

1986-01-01

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

Effects of residual stress on irradiation hardening in stainless steels

Energy Technology Data Exchange (ETDEWEB)

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

2007-07-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Ion irradiation-induced precipitation of Cr23C6 at dislocation loops in austenitic steel

International Nuclear Information System (INIS)

Jin, Shuoxue; Guo, Liping; Luo, Fengfeng; Yao, Zhongwen; Ma, Shuli; Tang, Rui

2013-01-01

The irradiation-induced precipitates in argon ion-irradiated austenitic stainless steel at 550 °C were examined via transmission electron microscopy. The selected-area electron diffraction patterns of precipitates indicated unambiguously that the precipitates were Cr 23 C 6 carbides. It was observed directly for the first time that irradiation-induced Cr 23 C 6 precipitates formed at dislocation loops in austenitic stainless steel, and coarsened with increasing irradiation dose.

Reference manual on the IAEA JRQ correlation monitor steel for irradiation damage studies

International Nuclear Information System (INIS)

2001-07-01

The objective of this report is to provide information on the mechanical properties of the ASTM A533 grade B class 1 steel that was designated as 'JRQ reference steel' and for many years served as a radiation/mechanical property correlation monitor in a number of international and national studies of irradiation embrittlement of reactor pressure vessel steel. This report provides the most comprehensive listing of material test data obtained on the JRQ manufacturing history and material properties in the initial, and as delivered condition during the implementation of two IAEA co-ordinated research projects (CRPs) on behaviour of reactor pressure vessel steels under neutron irradiation

Precipitation behavior in austenitic and ferritic steels during fast neutron irradiation and thermal aging*1

Science.gov (United States)

Kawanishi, H.; Hajima, R.; Sekimura, N.; Arai, Y.; Ishino, S.

1988-07-01

Precipitation behavior has been studied using a carbon extraction replica technique in Ti-modified Type 316 stainless steels (JPCA-2) and 9Cr-2Mo ferritic/martensitic steels (JFMS) irradiated to 8.1 × 10 24 n/m 2 at 873 and 673 K, respectively, in the experimental fast breeder reactor JOYO. Precipitate identification and compositional analysis were carried out on extracted replicas. The results were compared to those from the as-received steel and a control which had been given the same thermal as-treatment as the specimens received during irradiations. Carbides, Ti-sulphides and phosphides were precipitated in JPCA-2. Precipitate observed in JFMS included carbides, Laves-phases and phosphides. The precipitates in both steels were concluded to be stable under irradiation except for MC and M 6C in JPCA-2. Small MC particles were found precipitated in JPCA-2 during both irradiation and aging. Irradiation proved to promote the precipitation of M 6C in JPCA-2.

Microstructural stability of fast reactor irradiated 10 to 12% Cr ferritic-martensitic stainless steels

International Nuclear Information System (INIS)

Little, E.A.; Stoter, L.P.

1982-01-01

The strength and microstructural stability of three 10 to 12% Cr ferritic-martensitic stainless steels have been characterized following fast reactor irradiation to damage levels of 30 displacements per atom (dpa) at temperatures in the range 380 to 615 0 C. Irradiation results in either increases or decreases in room temperature hardness depending on the irradiation temperature. These strength changes can be qualitatively rationalized in terms of the combined effects of irradiation-induced interstitial dislocation loop formation and recovery of the dislocation networks comprising the initial tempered martensite structures. Precipitate evolution in the irradiated steels is associated with the nonequilibrium segregation of the elements nickel, silicon, molybdenum, chromium and phosphorus, brought about by solute-point defect interactions. The principal irradiation-induced precipitates identified are M 6 X, intermetallic chi and sigma phases and also α' (Cr-rich ferrite). The implications of the observed microstructural changes on the selection of martensitic stainless steels for fast reactor wrapper applications are briefly considered

Heavy-section steel irradiation program: Embrittlement issues

International Nuclear Information System (INIS)

Corwin, W.R.

1991-01-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

Irradiation Assisted Stress Corrosion Cracking of austenitic stainless steels

Energy Technology Data Exchange (ETDEWEB)

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

1998-03-01

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

Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

Microstructural evolution in reactor pressure vessel steel under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

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.

Pilot-scale ultrafiltration testing for the F and H area effluent treatment facility

International Nuclear Information System (INIS)

Kessler, J.L.

1984-01-01

An F and H Area Effluent Treatment Facility (F/H ETF) is being designed to treat low activity aqueous effluents which are produced from F and H Area daily operations. The treatment scheme for the F/H ETF will include pretreatment (pH adjustment and filtration) followed by Reverse Osmosis and/or Ion Exchange to remove dissolved species. Several alternative treatment processes are being considered for the F/H ETF. One of the alternatives in the pretreatment step is tubular Ultrafiltration (UF), using a dynamically formed zirconium oxide membrane supported on a porous stainless steel backing. Pilot-scale testing with a single membrane module (13 ft 2 area) and 200-Area effluent simulant has demonstrated that UF is a viable filtration option for the F/H ETF. UF testing at TNX has defined the operating conditions necessary for extended operation and also demonstrated excellent filtration performance (filtrate SDI 2 /day) flux and will provide excellent pretreatment for both reverse osmosis and ion exchange. 2 refs

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....

Heavy-section steel irradiation program. Progress report, October 1992--March 1993

International Nuclear Information System (INIS)

Corwin, W.R.

1998-04-01

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is one of only two more safety-related components of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established at Oak Ridge National Laboratory (ORNL) under sponsorship of the Nuclear Regulatory Commission (NRC). The primary goal of this major safety program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (in particular, the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program centers on experimental assessments of irradiation-induced embrittlement (including the completion of certain irradiation studies previously conducted by the Heavy-Section Steel Technology Program) augmented by detailed examinations and modeling of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties

Competitiveness values of irradiated adults of callosobruchus maculatus (F.) irradiated as mature pupae

International Nuclear Information System (INIS)

Ahmed, M.Y.Y.

1981-01-01

When mature pupae of Callosobruchus maculatus were treated with 3 Krad, the resulting adults were sterile when they were paired with untreated opposite sex. Males and females both treated with a sterilizing dose (3 Krad) and confined with untreated (U) males and females at a I male: I female: U male: U female (irradiated males: irradiated females: unirradiated males: unirradiated females) ratio caused 69.1% infertility in the resulting eggs. When the ratio of sterile males and females was increased to 5.5:1:1; 10:10:1:1 or 15:1:1 (I male: female: U male: U female) the percentage infertility reached 82.5, 95.0 and 100.0, respectively. The percentage of observed infertility was less than the expected infertility for the ratios 1:1:1:1:5:5:1:1 and 10:10:1:1, but it was exceeded with the highest ratio used (15:15:1:1). Competitiveness values for irradiated adults increased with an increasing ratio of irradiated to unirradiated adults. Since the ratio of 15:15:1:1 gave rise to 100% egg infertility (the expected infertility was 99.6%), no F 1 adults was produced; and the competitiveness value slightly exceeded 1.0 (i.e. the sterile adults were fully competitive with the normal ones). These results indicated that irradiation with 3 Krad, a sterilizing dose, did not decrease sexual competitiveness of irradiated adults. Also, the release of (I) females together with (I) males could give good results in controlling a population of C. maculatus in a autocidal control program; and, therefore, separation of the sexes prior to release is probably unnecessary. (author)

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

International Nuclear Information System (INIS)

Rouxel, Baptiste

2016-01-01

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

Irradiation creep of 316 and 316 Ti steels

International Nuclear Information System (INIS)

Lehmann, J.; Dupouy, J.M.; Boutard, J.L.; Maillard, A.; Broudeur, R.

1979-07-01

Irradiation creep results for several 316 and 316 Ti steels show the effects of stress, dose and temperature in the range 400 to 550 0 C. The observed differences are related to the compositioning and metallurgical conditions of the materials. (author)

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Effects of irradiation on the fracture behavior of austenitic stainless steels

International Nuclear Information System (INIS)

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

1977-01-01

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

Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

International Nuclear Information System (INIS)

Hirose, T.; Ogiwara, H.; Enoeda, M.; Akiba, M.

2007-01-01

Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 μm and 10 μm were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ogiwara, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Naka-gun, Ibaraki-ken (Japan); Akiba, M. [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki (Japan)

2007-07-01

Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 {mu}m and 10 {mu}m were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

International Nuclear Information System (INIS)

Brauer, G.; Matz, W.; Liszkay, L.; Molnar, B.

1990-11-01

Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

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

International Nuclear Information System (INIS)

Kohyama, Akira; Donomae, Takako

1999-03-01

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

Status report on experiments and modelling of the cleavage fracture behaviour of F82Hmod using local fracture grid. Task TTMS-005

International Nuclear Information System (INIS)

Riesch-Oppermann, H.; Walter, M.

2001-09-01

Within the European Fusion Technology Programme framework, a fracture mechanics description of the material behaviour in the ductile to brittle transition-regime is developed using local fracture criteria. Based on experimental results using axisymmetrically notched and pre-cracked specimens together with a numerical stress analysis at fracture load, a statistical evaluation of cleavage fracture parameters can be performed along the lines described in various code schemes such as the British Energy R6-Code or the ESIS P6 procedure. The report contains results of the experimental characterization of the deformation and fracture behaviour of the fusion candidate RAFM steel variant F82Hmod, details and background of the numerical procedure for cleavage fracture parameter determination as well as additional statistical inference methods for transferability analysis. Fractographic results give important information about fracture mode and fracture origin sites and their location. Numerical prediction of fracture origin distribution is an important tool for transferability assessment. Future issues comprise constraint effect and ductile damage as well as incorporation of irradiation effects, which are topically addressed. The methodology developed and described in the present report will be applied to characterize material behaviour of future RAFM variants as the EUROFER 97, for which analysis is currently under way. (orig.)

Microchemical evolution of neutron-irradiated stainless steel

International Nuclear Information System (INIS)

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

1980-04-01

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

Effect of microstructure on radiation induced segregation and depletion in ion irradiated SS316 steel

International Nuclear Information System (INIS)

Jin, Hyung Ha; Kwon, Sang Chul; Kwon, Jun Hyun

2011-01-01

Irradiation assisted stress corrosion cracking (IASCC), void swelling and irradiation induced hardening are caused by change of characteristics of material by neutron irradiation, stress state of material and environmental situation. It has been known that chemical compositions varies at grain boundary (GB) significantly with fluence level and the depletion of Cr element at GB has been considered as one of important factors causing material degradation, especially, IASCC in austenitic stainless steel. However, experimental results of IASCC under PWR condition were directly not connected with Cr depletion phenomenon by neutron irradiation. Because the mechanism of IASCC under PWR has not yet been clearly understood in spite of many energetic researches, fundamental researches about radiation induced segregation and depletion in irradiated austenitic stainless steels have been attracted again. In this work, an effect of residual microstructure on radiation induced segregation and depletion of alloy elements at GB was investigated in ion irradiated SS316 steel using transmission electron microscope (TEM) with energy dispersive spectrometer (EDS)

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

Energy Technology Data Exchange (ETDEWEB)

Hug, E., E-mail: eric.hug@ensicaen.fr [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Prasath Babu, R. [School of Materials, University of Manchester, M13 9PL (United Kingdom); Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Monnet, I. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Etienne, A. [Groupe de Physique des Matériaux, UMR CNRS 6634, Université et INSA de Rouen, Normandie Université, Saint-Etienne du Rouvray Cedex (France); Moisy, F. [Centre de recherches sur les Ions, les Matériaux et la Photonique CEA-CNRS, Normandie Université, 6 Bd Maréchal Juin, 14050 Caen (France); Pralong, V. [Laboratoire de Cristallographie et Sciences des Matériaux, Normandie Université, CNRS UMR 6508, 6 Bd Maréchal Juin, 14050 Caen (France); Enikeev, N. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); Saint Petersburg State University, Laboratory of the Mechanics of Bulk Nanostructured Materials, 198504 St. Petersburg (Russian Federation); Abramova, M. [Institute of Physics of Advanced Materials, Ufa (Russian Federation); and others

2017-01-15

Highlights: • Impacts of nanostructuration and irradiation on the properties of 316 stainless steels are reported. • Irradiation of nanostructured samples implies chromium depletion as than depicted in coarse grain specimens. • Hardness of nanocrystalline steels is only weakly affected by irradiation. • Corrosion resistance of the nanostructured and irradiated samples is less affected by the chromium depletion. - Abstract: 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 {sup 56}Fe{sup 5+} 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.

Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

International Nuclear Information System (INIS)

Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

2011-01-01

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

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

International Nuclear Information System (INIS)

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

1979-01-01

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

The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

International Nuclear Information System (INIS)

Jones, R.B.; Buswell, J.T.

1987-02-01

An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

Influence of He ions irradiation on the deuterium permeation and retention behavior in the CLF-1 steel

International Nuclear Information System (INIS)

Xu, Yu-Ping; Lu, Tao; Li, Xiao-Chun; Liu, Feng; Liu, Hao-Dong; Wang, Jing; An, Zhong-Qing; Ding, Fang; Hong, Suk-Ho; Zhou, Hai-Shan; Luo, Guang-Nan

2016-01-01

To evaluate the influence of He ions irradiation on the deuterium permeation and retention behavior in RAFM steels, samples made of the CLF-1 steel was irradiated with 3.5 MeV He ions. Gas driven permeation experiments were performed, and the permeability of virgin sample and pre-irradiated sample were obtained and compared. In order to characterize the effect of He ions irradiation on the deuterium retention behavior, deuterium gas exposure was carried out at 623 K, followed by thermal desorption spectra experiments. The total deuterium retention of the CLF-1 steel increased owing to He ions implantation, which could be attributed to the increase in trapping site for deuterium by the He pre-irradiation.

Resonant creep enhancement in austenitic stainless steels due to pulsed irradiation at low doses

International Nuclear Information System (INIS)

Kishimoto, N.; Amekura, H.; Saito, T.

1994-01-01

Steady-state irradiation creep of austenitic stainless steels has been extensively studied as one of the most important design parameters in fusion reactors. The steady-state irradiation creep has been evaluated using in-pile and light-ion experiments. Those creep compliances of various austenitic steels range in the vicinity of ε/Gσ = 10 -6 ∼10 -5 (dpa sm-bullet MPa) -1 , depending on chemical composition etc. The mechanism of steady-state irradiation creep has been elucidated, essentially in terms of stress-induced preferential absorption of point defects into dislocations, and their climb motion. From this standpoint, low doses such as 10 -3 ∼10 -1 dpa would not give rise to any serious creep, and the irradiation creep may not be a critical issue for the low-dose fusion devices including ITER. It is, however, possible that pulsed irradiation causes different creep behaviors from the steady-state one due to dynamic unbalance of interstitials and vacancies. The authors have actually observed anomalous creep enhancement due to pulsed irradiation in austenitic stainless steels. The resonant behavior of creep indicates that pulsed irradiation may cause significant deformation in austenitic steels even at such low doses and slow pulsing rates, especially for the SA-materials. The first-wall materials in plasma operation of ∼10 2 s may suffer from unexpected transient creep, even in the near-term fusion deices, such as ITER. Though this effect might be a transient effect for a relatively short period, it should be taken into account that the pulsed irradiation makes influences on stress relaxation of the fusion components and on the irradiation fatigue. The mechanism and the relevant behaviors of pulse-induced creep will be discussed in terms of a point-defect model based on the resonant interstitial enrichment

APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Blavette, D.

1996-01-01

Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

Corrosion behavior of steels in liquid lead bismuth with low oxygen concentrations

Science.gov (United States)

Kurata, Yuji; Futakawa, Masatoshi; Saito, Shigeru

2008-02-01

Corrosion tests in pots were conducted to elucidate corrosion behavior of various steels in liquid lead-bismuth for 3000 h under the condition of an oxygen concentration of 5 × 10 -8 wt% at 450 °C and an oxygen concentration of 3 × 10 -9 wt% at 550 °C, respectively. Significant corrosion was not observed at 450 °C for ferritic/martensitic steels, F82H, Mod.9Cr-1Mo steel, 410SS, 430SS except 2.25Cr-1Mo steel. Pb-Bi penetration into steels and dissolution of elements into Pb-Bi were severe at 550 °C even for ferritic/martensitic steels. Typical dissolution attack occurred for pure iron both at 550 °C without surface Fe 3O 4 and at 450 °C with a thin Fe 3O 4 film. Ferritization due to dissolution of Ni and Cr, and Pb-Bi penetration were recognized for austenitic stainless steels, 316SS and 14Cr-16Ni-2Mo steel at both temperatures of 450 °C and 550 °C. The phenomena were mitigated for 18Cr-20Ni-5Si steel. In some cases oxide films could not be a corrosion barrier in liquid lead-bismuth.

Irradiation effects in strain aged pressure vessel steel

Energy Technology Data Exchange (ETDEWEB)

Grounes, M; Myers, H P

1962-02-15

Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

Evaluation of strain-rate sensitivity of ion-irradiated austenitic steel using strain-rate jump nanoindentation tests

Energy Technology Data Exchange (ETDEWEB)

Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University Gokasho, Uji 611-0011, Kyoto (Japan); Hamaguchi, Dai; Ando, Masami; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan)

2016-11-01

Highlights: • We examined strain-rate jump nanoindentation on ion-irradiated stainless steel. • We observed irradiation hardening of the ion-irradiated stainless steel. • We found that strain-rate sensitivity parameter was slightly decreased after the ion-irradiation. - Abstract: The present study investigated strain-rate sensitivity (SRS) of a single crystal Fe–15Cr–20Ni austenitic steel before and after 10.5 MeV Fe{sup 3+} ion-irradiation up to 10 dpa at 300 °C using a strain-rate jump (SRJ) nanoindentation test. It was found that the SRJ nanoindentation test is suitable for evaluating the SRS at strain-rates from 0.001 to 0.2 s{sup −1}. Indentation size effect was observed for depth dependence of nanoindentation hardness but not the SRS. The ion-irradiation increased the hardness at the shallow depth region but decreased the SRS slightly.

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

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)

Effect of Proton Irradiation on the Corrosion Behaviors of Ferritic/Martensitic Steel in Liquid Metal Environment

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeonghyeon; Kim, Tae Yong; Kim, Ji Hyun [UNIST, Ulsan (Korea, Republic of)

2016-10-15

Liquid metal fast breeder reactors (LMFBRs) such as sodium-cooled fast reactor (SFR) and lead-cooled fast reactor (LFR) are the candidates of GEN-IV nuclear energy systems. Among various liquid metals that can be used as primary coolant material, sodium is a world widely used coolant for GEN-IV reactors. In this study, as-received Gr.92 and irradiated Gr.92 specimen in the oxygen-saturated liquid sodium were examined at high temperature for 300h. The microstructure results reveal the information of the effect of irradiation and effect of the chrome concentration in specimen. From the SRIM result, penetration distance of 40 μm in stainless steel and nominal sample thickness of 30 μm was used to avoid the damage peak and any proton implantation and From the microstructural evaluation, chromium-rich zones existed under the surface of the both of non-irradiated and irradiated materials. The irradiated materials showed chromium-rich zones with larger depths than the non-irradiated specimens.

Depth profiling of hydrogen in ferritic/martensitic steels by means of a tritium imaging plate technique

International Nuclear Information System (INIS)

Otsuka, Teppei; Tanabe, Tetsuo

2013-01-01

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

Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

International Nuclear Information System (INIS)

Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

1998-01-01

One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

Low temperature fatigue crack propagation in neutron irradiated Type 316 steel and weld metal

International Nuclear Information System (INIS)

Lloyd, G.J.; Walls, J.D.; Gravenor, J.

1981-02-01

The fast cycling fatigue crack propagation characteristics of Type 316 steel and weld metal have been investigated at 380 0 C after irradiation to 1.72-1.92x10 20 n/cm 2 (E>1MeV) and 2.03x10 21 n/cm 2 (E>1MeV) at the same temperature. With mill-annealed Type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked Type 316 steel irradiation to 2.03x10 21 n/cm 2 (E>1MeV) caused increases in the rate of crack propagation. For Type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered. The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation together with the results of control tensile experiments made on similarly irradiated materials. (author)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Microstructural probing of ferritic/martensitic steels using internal transmutation-based positron source

Energy Technology Data Exchange (ETDEWEB)

Krsjak, Vladimir, E-mail: vladimir.krsjak@psi.ch; Dai, Yong

2015-10-15

This paper presents the use of an internal {sup 44}Ti/{sup 44}Sc radioisotope source for a direct microstructural characterization of ferritic/martensitic (f/m) steels after irradiation in targets of spallation neutron sources. Gamma spectroscopy measurements show a production of ∼1MBq of {sup 44}Ti per 1 g of f/m steels irradiated at 1 dpa (displaced per atom) in the mixed proton–neutron spectrum at the Swiss spallation neutron source (SINQ). In the decay chain {sup 44}Ti → {sup 44}Sc → {sup 44}Ca, positrons are produced together with prompt gamma rays which enable the application of different positron annihilation spectroscopy (PAS) analyses, including lifetime and Doppler broadening spectroscopy. Due to the high production yield, long half-life and relatively high energy of positrons of {sup 44}Ti, this methodology opens up new potential for simple, effective and inexpensive characterization of radiation induced defects in f/m steels irradiated in a spallation target.

Microstructural probing of ferritic/martensitic steels using internal transmutation-based positron source

Science.gov (United States)

Krsjak, Vladimir; Dai, Yong

2015-10-01

This paper presents the use of an internal 44Ti/44Sc radioisotope source for a direct microstructural characterization of ferritic/martensitic (f/m) steels after irradiation in targets of spallation neutron sources. Gamma spectroscopy measurements show a production of ∼1MBq of 44Ti per 1 g of f/m steels irradiated at 1 dpa (displaced per atom) in the mixed proton-neutron spectrum at the Swiss spallation neutron source (SINQ). In the decay chain 44Ti → 44Sc → 44Ca, positrons are produced together with prompt gamma rays which enable the application of different positron annihilation spectroscopy (PAS) analyses, including lifetime and Doppler broadening spectroscopy. Due to the high production yield, long half-life and relatively high energy of positrons of 44Ti, this methodology opens up new potential for simple, effective and inexpensive characterization of radiation induced defects in f/m steels irradiated in a spallation target.

Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

International Nuclear Information System (INIS)

Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

2012-01-01

The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

Atom probe tomography of the evolution of the nanostructure of oxide dispersion strengthened steels under ion irradiation

Science.gov (United States)

Orlov, N. N.; Rogozhkin, S. V.; Bogachev, A. A.; Korchuganova, O. A.; Nikitin, A. A.; Zaluzhnyi, A. G.; Kozodaev, M. A.; Kulevoy, T. V.; Kuibeda, R. P.; Fedin, P. A.; Chalykh, B. B.; Lindau, R.; Hoffmann, Ya.; Möslang, A.; Vladimirov, P.

2017-09-01

The atom probe tomography of the nanostructure evolution in ODS1 Eurofer, ODS 13.5Cr, and ODS 13.5Cr-0.3Ti steels under heavy ion irradiation at 300 and 573 K is performed. The samples were irradiated by 5.6 MeV Fe2+ ions and 4.8 MeV Ti2+ ions to a fluence of 1015 cm-2. It is shown that the number of nanoclusters increases by a factor of 2-3 after irradiation. The chemical composition of the clusters in the steels changes after irradiation at 300 K, whereas the chemical composition of the clusters in the 13.5Cr-0.3Ti ODS steel remains the same after irradiation at 573 K.

Void shrinkage in stainless steel during high energy electron irradiation

International Nuclear Information System (INIS)

Singh, B.N.; Foreman, A.J.E.

1976-03-01

During irradiation of thin foils of an austenitic stainless steel in a high voltage electron microscope, steadily growing voids have been observed to suddenly shrink and disappear at the irradiation temperature of 650 0 Cthe phenomenon has been observed in specimens both with and withoutimplanted helium. Possible mechanisms for void shrinkage during irradiation are considered. It is suggested that the dislocation-pipe-diffusion of vacancies from or of self-interstitial atoms to the voids can explain the shrinkage behaviour of voids observed during our experiments. (author)

Optimized chemical composition, working and heat treatment condition for resistance to irradiation assisted stress corrosion cracking of cold worked 316 and high-chromium austenitic stainless steel

International Nuclear Information System (INIS)

Yonezawa, Toshio; Iwamura, Toshihiko; Fujimoto, Koji; Ajiki, Kazuhide

2000-01-01

The authors have reported that the primary water stress corrosion cracking (PWSCC) in baffle former bolts made of austenitic stainless steels for PWR after long-term operation is caused by irradiation-induced grain boundary segregation. The resistance to PWSCC of simulated austenitic stainless steels whose chemical compositions are simulated to the grain boundary chemical composition of 316 stainless steel after irradiation increased with decrease of the silicon content, increases of the chromium content, and precipitation of M 23 C 6 carbides at the grain boundaries. In order to develop resistance to irradiation assisted stress corrosion cracking in austenitic stainless steels, optimized chemical compositions and heat treatment conditions for 316CW and high-chromium austenitic stainless steels for PWR baffle former bolts were investigated. For 316CW stainless steel, ultra-low-impurities and high-chromium content are beneficial. About 20% cold working before aging and after solution treatment has also been recommended to recover sensitization and make M 23 C 6 carbides coherent with the matrix at the grain boundaries. Heating at 700 to 725degC for 20 to 50 h was selected as a suitable aging procedure. Cold working of 5 to 10% after aging produced the required mechanical properties. The optimized composition of the high-chromium austenitic stainless steel contents 30% chromium, 30% nickel, and ultra-low impurity levels. This composition also reduces the difference between its thermal expansion coefficient and that of 304 stainless steel for baffle plates. Aging at 700 to 725degC for longer than 40 h and cold working of 10 to 15% after aging were selected to meet mechanical property specifications. (author)

Improvement of the mechanical and frictional properties of steels by continuous and pulsed ion irradiation

International Nuclear Information System (INIS)

Romanov, I.G.

1992-01-01

Effect of continuous and powerful pulsed ion beams (PIB) on structural, mechanical, tribological properties and surface morphology of steels were investigated. The results obtained demonstrate the significant influence of ion irradiation type on microhardness, friction coefficient, wear resistance and surface roughness characteristics. Friction coefficient variation in irradiated steels is interpreted within the framework of an adhesion-deformation model

The natural aging of austenitic stainless steels irradiated with fast neutrons

Science.gov (United States)

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

2018-02-01

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

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

International Nuclear Information System (INIS)

Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

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)

Characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-12-01

In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

Energy Technology Data Exchange (ETDEWEB)

Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

2017-05-15

Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

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

International Nuclear Information System (INIS)

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

1980-01-01

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

The radiation swelling effect on fracture properties and fracture mechanisms of irradiated austenitic steels. Part I. Ductility and fracture toughness

Energy Technology Data Exchange (ETDEWEB)

Margolin, B., E-mail: mail@crism.ru; Sorokin, A.; Shvetsova, V.; Minkin, A.; Potapova, V.; Smirnov, V.

2016-11-15

The radiation swelling effect on the fracture properties of irradiated austenitic steels under static loading has been studied and analyzed from the mechanical and physical viewpoints. Experimental data on the stress-strain curves, fracture strain, fracture toughness and fracture mechanisms have been represented for austenitic steel of 18Cr-10Ni-Ti grade (Russian analog of AISI 321 steel) irradiated up to neutron dose of 150 dpa with various swelling. Some phenomena in mechanical behaviour of irradiated austenitic steels have been revealed and explained as follows: a sharp decrease of fracture toughness with swelling growth; untypical large increase of fracture toughness with decrease of the test temperature; some increase of fracture toughness after preliminary cyclic loading. Role of channel deformation and channel fracture has been clarified in the properties of irradiated austenitic steel and different tendencies to channel deformation have been shown and explained for the same austenitic steel irradiated at different temperatures and neutron doses.

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.

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.

Neutron irradiation test of copper alloy/stainless steel joint materials

International Nuclear Information System (INIS)

Yamada, Hirokazu; Kawamura, Hiroshi

2006-01-01

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

Deformation mechanisms in ferritic/martensitic steels and the impact on mechanical design

International Nuclear Information System (INIS)

Ghoniem, Nasr M.; Po, Giacomo; Sharafat, Shahram

2013-01-01

Structural steels for nuclear applications have undergone rapid development during the past few decades, thanks to a combination of trial-and-error, mechanism-based optimization, and multiscale modeling approaches. Deformation mechanisms are shown to be intimately related to mechanical design via dominant plastic deformation modes. Because mechanical design rules are mostly based on failure modes associated with plastic strain damage accumulation, we present here the fundamental deformation mechanisms for Ferritic/Martensitic (F/M) steels, and delineate their operational range of temperature and stress. The connection between deformation mechanisms, failure modes, and mechanical design is shown through application of design rules. A specific example is given for the alloy F82H utilized in the design of a Test Blanket Module (TBM) in the International Thermonuclear Experimental Reactor (ITER), where several constitutive equations are developed for design-related mechanical properties

Deformation mechanisms in ferritic/martensitic steels and the impact on mechanical design

Energy Technology Data Exchange (ETDEWEB)

Ghoniem, Nasr M., E-mail: ghoniem@seas.ucla.edu; Po, Giacomo; Sharafat, Shahram

2013-10-15

Structural steels for nuclear applications have undergone rapid development during the past few decades, thanks to a combination of trial-and-error, mechanism-based optimization, and multiscale modeling approaches. Deformation mechanisms are shown to be intimately related to mechanical design via dominant plastic deformation modes. Because mechanical design rules are mostly based on failure modes associated with plastic strain damage accumulation, we present here the fundamental deformation mechanisms for Ferritic/Martensitic (F/M) steels, and delineate their operational range of temperature and stress. The connection between deformation mechanisms, failure modes, and mechanical design is shown through application of design rules. A specific example is given for the alloy F82H utilized in the design of a Test Blanket Module (TBM) in the International Thermonuclear Experimental Reactor (ITER), where several constitutive equations are developed for design-related mechanical properties.

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.

Correlation between the thermoluminescence and the F centre thermal stability in KBr, KI, NaCl and NaF irradiated at room temperature

International Nuclear Information System (INIS)

Mariani, D. F.

1977-01-01

Over the years several models have been proposed to explain the thermoluminescence in irradiated alkali halides. To verify these models, the thermoluminescence and the thermal stability of the F centres in KBr, K l, NaCl and NaF single crystals irradiated at room temperature with x or gamma rays have been studied. It has been concluded that both phenomena, the thermoluminescence and the F centre thermal stability are two different aspects of the same basic phenomenon, i.e. the recombination of mobile interstitials thermally released from traps with F centres (Ausin V. and Alvarez Rivas J.L. 1972). It has been found in some cases, that the H centres also play the role of recombination centres for interstitials. (Author) 42 refs

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

International Nuclear Information System (INIS)

Klueh, R.L.

1997-01-01

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

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)

Heavy-Section Steel Irradiation Program. Volume 5, No. 2, Progress report, April 1994--September 1994

International Nuclear Information System (INIS)

Corwin, W.R.

1995-07-01

The Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness curve shift in high-copper weldments (Series 5 and 6), (3) K lc and K la curve shifts in low upper-shelf (LUS) welds (Series 8), (4) irradiation effects in a commercial LUS weld (Series 10), (5) irradiation effects on weld heat-affected zone and plate materials (Series 11), (6) annealing effects in LUS welds (Series 9), (7) microstructural and microfracture analysis of irradiation effects, (8) in-service irradiated and aged material evaluations, (9) Japan Power Development Reactor (JPDR) steel examination, (10) fracture toughness curve shift method, (11) special technical assistance, (12) technical assistance for Joint Coordinating Committee on Civilian Nuclear Reactor Safety (JCCCNRS) Working Groups 3 and 12, (13) correlation monitor materials, and (14) test reactor coordination. Progress on each task is reported

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

International Nuclear Information System (INIS)

Corwin, W.R.

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 Ic ) curve shift in high-copper welds, (3) crack-arrest toughness (K Ia ) curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K 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

Heavy-Section Steel Irradiation Program: Volume 3, Progress report, October 1991--September 1992

International Nuclear Information System (INIS)

Corwin, W.R.

1995-02-01

The primary goal of the Heavy-Section Steel Irradiation Program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 10 tasks: (1) program management, (2) K Ic curve shift in high-copper welds, (3) K Ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K Ia curve shifts in low upper-shelf welds, (6) irradiation effects in a commercial low upper-shelf 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 October 1991 to September 1992

Radiation response of ODS ferritic steels with different oxide particles under ion-irradiation at 550 °C

Science.gov (United States)

Song, Peng; Morrall, Daniel; Zhang, Zhexian; Yabuuchi, Kiyohiro; Kimura, Akihiko

2018-04-01

In order to investigate the effects of oxide particles on radiation response such as hardness change and microstructural evolution, three types of oxide dispersion strengthened (ODS) ferritic steels (named Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS), mostly strengthened by Y-Ti-O, Y-Al-O and Y-Zr-O dispersoids, respectively, were simultaneously irradiated with iron and helium ions at 550 °C up to a damage of 30 dpa and a corresponding helium (He) concentration of ∼3500 appm to a depth of 1000-1300 nm. A single iron ion beam irradiation was also performed for reference. Transmission electron microscopy revealed that after the dual ion irradiation helium bubbles of 2.8, 6.6 and 4.5 nm in mean diameter with the corresponding number densities of 1.1 × 1023, 2.7 × 1022 and 3.6 × 1022 m-3 were observed in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS, respectively, while no such bubbles were observed after single ion irradiation. About 80% of intragranular He bubbles were adjacent to oxide particles in the ODS ferritic steels. Although the high number density He bubbles were observed in the ODS steels, the void swelling in Y-Ti-ODS, Y-Al-ODS and Y-Al-Zr-ODS was still small and estimated to be 0.13%, 0.53% and 0.20%, respectively. The excellent swelling resistance is dominantly attributed to the high sink strength of oxide particles that depends on the morphology of particle dispersion rather than the crystal structure of the particles. In contrast, no dislocation loops were produced in any of the irradiated steels. Nanoindentation measurements showed that no irradiation hardening but softening was found in the ODS ferritic steels, which was probably due to irradiation induced dislocation recovery. The helium bubbles in high number density never contributed to the irradiation hardening of the ODS steels at these irradiation conditions.

Irradiation hardening of Mod.9Cr-1Mo steel

International Nuclear Information System (INIS)

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

2009-01-01

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

Mechanical properties and TEM examination of RAFM steels irradiated up to 70 dpa in BOR-60

Energy Technology Data Exchange (ETDEWEB)

Gaganidze, E., E-mail: Ermile.Gaganidze@kit.edu [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Petersen, C.; Materna-Morris, E.; Dethloff, C.; Weiss, O.J.; Aktaa, J. [Karlsruher Institut fuer Technologie, Institut fuer Angewandte Materialien, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Povstyanko, A.; Fedoseev, A.; Makarov, O.; Prokhorov, V. [Joint Stock Company ' State Scientific Centre Research Institute of Atomic Reactors' , 433510 Dimitrovgrad-10, Ulyanovsk Region (Russian Federation)

2011-10-01

Mechanical properties of Reduced Activation Ferritic/Martensitic (RAFM) steels were studied after irradiation in BOR-60 reactor to a neutron displacement damage of 70 dpa at 330-340 deg. C. Yield stress and Ductile-to-Brittle-Transition-Temperature of EUROFER97 indicate saturation of hardening and embrittlement. The phenomenological models for description of microstructure evolution and resulting irradiation hardening and embrittlement are discussed. The evolution of yield stress with dose is qualitatively understood within a Whapham and Makin model. Dislocation loops examined in TEM are considered a main source for low-temperature irradiation hardening. The analysis of the fatigue data in terms of the inelastic strain reveals comparable fatigue behaviour both for unirradiated and irradiated conditions, which can be described by a common Manson-Coffin relation. The study of helium effects in B-doped model steels indicated progressive material embrittlement with helium content. Post-irradiation annealing of RAFM steels yielded substantial recovery of mechanical properties.

Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

International Nuclear Information System (INIS)

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

The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

Energy Technology Data Exchange (ETDEWEB)

Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

1994-12-31

Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

International Nuclear Information System (INIS)

Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

2014-01-01

The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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.)

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.

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

Science.gov (United States)

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

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

Behavior of implanted hydrogen in ferritic/martensitic steels under irradiation

Science.gov (United States)

Wan, F.; Takahashi, H.; Ohnuki, S.; Nagasaki, R.

1988-07-01

The aim of this study was to clarify the behavior of hydrogen under irradiation in ferritic/martensitic stainless steel Fe-10Cr-2Mo-1Ni. Hydrogen was implanted into the specimens by ion accelerator or chemical cathodic charging method, followed by electron irradiation in a HVEM at temperatures from room temperature to 773 K. Streaks in the electron diffraction patterns were observed only during electron irradiation at 623-723 K. From these results it is suggested that the occurrence of the streak pattern is due to the formation of radiation-induced complexes of Ni or Cr with hydrogen along directions.

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

International Nuclear Information System (INIS)

Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

1993-01-01

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

Energy Technology Data Exchange (ETDEWEB)

Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

1994-12-31

In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

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

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

International Nuclear Information System (INIS)

Dethloff, Christian; Gaganidze, Ermile; Svetukhin, Vyacheslav V.; Aktaa, Jarir

2012-01-01

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 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.

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.