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Journal Articles

Corrosion behavior of Al-alloying high Cr-ODS steels in lead-bismuth eutectic

Takaya, Shigeru; Furukawa, Tomohiro; Aoto, Kazumi; M$"u$ller, G.*; Weisenburger, A.*; Heinzel, A.*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; et al.

Journal of Nuclear Materials, 386-388, p.507 - 510, 2009/04

 Times Cited Count:44 Percentile:95.19(Materials Science, Multidisciplinary)

The corrosion resistance of ODS steels with 0$$sim$$3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr and of a 12Cr steel were examined. The experiments were conducted at 550 and 650 $$^{circ}$$C up to 3,000 h in stagnant LBE containing 10$$^{-6}$$ and 10$$^{-8} $$wt% oxygen for the ODS steels and at 550 $$^{circ}$$C up to 5,000 h in stagnant LBE containing 10$$^{-8}$$ wt% oxygen for the 12Cr steel, respectively. Protective Al oxide scales were formed on the surfaces of ODS steels with about 3.5 wt% Al and 13.7$$sim$$17.3 wt% Cr. The addition of Al is very effective to improve the corrosion resistance of ODS steels. The ODS steel with 16 wt% Cr and no Al does not show any corrosion resistance except for the specimen exposed to LBE with 10$$^{-6}$$ wt% oxygen at 650 $$^{circ}$$C. It is not expected to improve the corrosion resistance by increasing solely Cr content.

Journal Articles

Latest design of liquid lithium target in IFMIF

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Cevolani, S.*; Chida, Teruo*; Ciotti, M.*; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; et al.

Fusion Engineering and Design, 83(7-9), p.1007 - 1014, 2008/12

 Times Cited Count:14 Percentile:70.13(Nuclear Science & Technology)

This paper describes the latest design of liquid lithium target system in IFMIF. Design requirement of the Li target is to provide a stable Li jet with a speed of 20 m/s to handle an averaged heat flux of 1 GW/m$$^{2}$$. A double reducer nozzle and a concaved flow are applied to the target design. On Li purification, a cold trap and two kinds of hot trap are applied to control impurities below permissible levels. Nitrogen concentration shall be controlled below 10 wppm by one of the hot trap. Tritium concentration shall be controlled below 1 wppm by an yttrium hot trap. To maintain reliable continuous operation, various diagnostics are attached to the target assembly. Among the target assembly, a back-plate made of RAFM is located in the most severe region of neutron irradiation (50 dpa/y). Therefore, two design options of replaceable back wall and their remote handling systems are under investigation.

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications (2) -Part I. Stability of Oxide Layer Formed on High-Chromium Steels in LBE under Oxygen Content and Temperature Fluctuation-

M$"u$ller, G.*; Schumacher, G.*; Heinzel, A.*; Weisenburger, A.*; Zimmermann, F.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2005-021, 33 Pages, 2005/08

JNC-TY9400-2005-021.pdf:2.69MB

The behaviour of protective oxide layers on P122 and its welds and of ODS steel is examined under conditions of fluctuating temperatures and oxygen concentrations to simulate hot spot and loss of oxygen control. P122 steel (12Cr) and its welded joints are exposed to LBE at 550$$^{circ}$$C for 4000h with oxygen concentrations changing between 10$$^{-6}$$ and 10$$^{-8}$$wt% every 800h. It is found that like in case of constant oxygen concentration of 10$$^{-6}$$wt% a protective spinel layer was maintained on P122 and also on its welded joint. Experiments with fluctuating temperatures from 550$$^{circ}$$C to 650$$^{circ}$$C and back every 800h yield satisfying corrosion protection up to 4800h only for the GESA treated ODS steel in LBE with 10$$^{-6}$$wt%. Original ODS steel in LBE with 10$$^{-6}$$ and 10$$^{-8}$$wt% oxygen showed local strong dissolution attack after 4800h in agreement with the behaviour of ODS in LBE with a constant temperature of 650$$^{circ}$$C. GESA treated ODS steel does not form stable protective alumina layers under condition of fluctuating temperatures in LBE with 10$$^{-8}$$wt% oxygen.

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications; Part VI. Results of Exposure Experiments in Oxygen Containing Flowing LBE at 550$$^{circ}$$C for 10,000h

Schroer, C.*; Voss, Z.*; Wedemeyer, O.*; Novotny, J.*; Konys, J.*; Heinzel, A.*; Weisenburger, A.*; M$"u$ller, G.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2005-020, 45 Pages, 2005/08

JNC-TY9400-2005-020.pdf:5.26MB

This report summarises the results of exposure experiments on the behaviour of 12Cr-2W steel P122 and an 9Cr-2W ODS-steel in oxygen-containing flowing lead-bismuth eutectic (LBE) at 550$$^{circ}$$C, which were performed in the CORRIDA loop at the Karlsruhe Lead Laboratory (KALLA) as part of the collaboration between the Japan Nuclear Cycle Development Institute and the Karlsruhe Research Centre. The duration of these experiments was nominally 800, 2000, 5000 and 10,000 h. Both steels were tested after surface-finishing by turning and after surface alloying with aluminium in the GESA-facility. Owing to initial problems with the enrichment of oxygen in the flowing LBE, the oxygen content considerably varied during most of these experiments, so that the influence of temporary changes in the oxygen content of the LBE could also be investigated. The behaviour of P122 and ODS at permanently high oxygen content was deduced from an experiment for 4990 h at 550$$^{circ}$$C and (c$$_{O}$$ $$approx$$ 5$$times$$10$$^{-7}$$ mass-% or a$$_{PbO}$$ $$approx$$ 10$$^{-3}$$). The results are compared with the findings of exposures to stagnant LBE at 550$$^{circ}$$C and c$$_{O}$$=10$$^{-6}$$ mass-% (COSTA-experiments).

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications (2); Part II. Exposure of Weld P122 in Oxygen Containing Flowing LBE at 550$$^{circ}$$C for 5,000h

Schroer, C.*; Voss, Z.*; Wedemeyer, O.*; Novotny, J.*; Konys, J.*; Heinzel, A.*; Weisenburger, A.*; M$"u$ller, G.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2005-019, 26 Pages, 2005/08

JNC-TY9400-2005-019.pdf:5.04MB

This report summarises the results of exposure experiments on the behaviour of 12Cr-2W steel P122 joined by multi-run TIG welding in oxygen-containing flowing lead-bismuth eutectic (LBE) at 550$$^{circ}$$C for nominally 800, 2000 and 5000 h. The flow velocity of the LBE and the mean oxygen content were 2 m/s and 5$$times$$10-7 mass-% (aPbO = 10-3), respectively. The influence of surface-alloying with aluminium (GESA-treatment) was also investigated in an exposure experiment for nominally 5000 h. The behaviour of the welded joint is generally comparable to that of P122 (after the standard heat-treatment), both qualitatively and quantitatively. Only few exceptions were observed which probably result from local peculiarities of the specific sample material. After surface alloying with aluminium, no significant oxidation and no liquid metal corrosion occurred in the centre of the specimens (including the weld seam), where the desired high quality of the aluminised layer was achieved.

Journal Articles

Stability of oxide layer formed on high-chromium steel in LBE under oxygen content and temperature fluctuation

Weisenburger, A.*; Aoto, Kazumi; M$"u$ller, G.*; Heinzel, A.*; Furukawa, Tomohiro

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13) (CD-ROM), 6 Pages, 2005/06

The behaviour of protective oxide layers on P122 and its welds and of ODS steel in LBE is examined under conditions of changing temperatures and oxygen concentrations. P122 teel ($$^{12}$$Cr) and its welded joints are exposed to LBE at 550 $$^{circ}$$C for 4.000 h with oxygen concentrations of 10$$^{-6}$$ and 10$$^{-8} $$wt% which change every 800 h. It is found that like in case f constant oxygen concentration of 10$$^{-6}$$ wt% a protective spinel layer was maintained on P122 and also on its welded joint. Two experiments are conducted on ODS steel, both with temperatures changing from 550 to 650 $$^{circ}$$C and back every 800 h, one experiment with 10$$^{-6}$$ the other with 10$$^{-8}$$ wt% oxygen in LBE. Like in the former test with constant emperature at 550 $$^{circ}$$C no dissolution attack could be observed in experiments with temperature fluctuation. Contrary to this results is the observed dissolution attack on ODS with a onstant temperature of 650 $$^{circ}$$C at 10$$^{-6}$$ wt% oxygen in which formation of a protective layer was not allowed before reaching 650 $$^{circ}$$C LBE temperature.

Journal Articles

Effect of oxygen concentration and temperature on compatibility of ODS steel with liquid, Stagnant Pb$$_{45}$$Bi$$_{55}$$

Furukawa, Tomohiro; M$"u$ller, G.*; Schumacher, G.*; Weisenburger, A.*; Heinzel, A.*; Aoto, Kazumi

Journal of Nuclear Materials, 335(2), p.189 - 193, 2004/11

In order to investigate the effect of oxygen concentration in LBE on corrosion behavior, exposure test of an oxide dispersion strength martensitic type steel was performed in stagnant lead-bismuth eutectic (LBE) containing 10$$^{-4}$$, 10$$^{-6}$$ and 10$$^{-8}$$wt% of oxygen at 500-650$$^{circ}$$C up to 10,000 hours. Results of metallurgical analysis, it was obtained that the base metal was protected from liquid metal corrosion by the formation of a spinel layer under the test condition containing 10$$^{-6}$$wt% of oxygen at 550$$^{circ}$$C or less. In LBE over 600$$^{circ}$$C containing 10$$^{-6}$$ and 10$$^{-4}$$wt% of oxygen, liquid metal corrosion was observed at some places on the surface. At the condition of 650$$^{circ}$$C containing 10$$^{-8}$$wt% of oxygen, although no oxide layer was observed on the surface, the base metal showed good compatibility with LBE.

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications; Part V. Results of Exposure Experiments in Oxygen Containing Flowing LBE at 550$$^{circ}$$C for 800 and 2000h

Schroer, G.*; Voss, V.*; Wedemeyer, O.*; Novotny, J.*; Konys, J.*; Heinzel, A.*; Weisenburger, A.*; M$"u$ller, G.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2004-023, 37 Pages, 2004/05

JNC-TY9400-2004-023.pdf:12.34MB

Journal Articles

Corrosion Behavior of FBR Candidate Materials in Stagnant Pb-Bi at Elevated Temperature Corrosion Behavior of FBR Candidate Materials in Stagnant Pb-Bi at Elevated Temperature

Furukawa, Tomohiro; M$"u$ller, G.*; Schumacher, G.*; Weisenburger, A.*; Heinzel, A.*; 2 of others*

Journal of Nuclear Science and Technology, 41(3), p.265 - 270, 2004/03

 Times Cited Count:42 Percentile:92.95(Nuclear Science & Technology)

Corrosion tests of the three Japanese materials named 316FR, 12 Cr-steel and ODS-M were performed in stagnant lead bismuth eutectic (LBE) of 500 - 650 degress centigrade for 5,000 h under oxygen control at 10$$^{-2}$$weight percent, and the corrosion prope

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications, 4; Corrosion investigation of steels at 550 and 650$$^{circ}$$C after 800, 2,000 and 5,000 h of exposure to stagnant liquid Pb-Bi containing 10$$^{-4}$$ and 10$$^{-8}$$ wt% of oxygen

Muller, G.*; Schumacher, G.*; Weisenburger, A.*; Heinzel, A.*; Zimmermann, F.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2003-028, 48 Pages, 2004/01

JNC-TY9400-2003-028.pdf:4.51MB

This is the fourth report on the compatibility of structural and fuel cladding materials that has to be investigated for a possible advanced heavy metal cooled reactor system. The first three reports considered the behavior of 316FR, P122 and ODS steels during 800, 2,000, 5,000 and 10,000 h in stagnant LBE between 500 and 650 degrees centigrade containing 10 sup-6 wt% of oxygen. This report describes experiments of 800, 2,000 and 5,000 h duration in liquid Pb-Bi containing 10 sup-4 and 10 sup-8 wt% of oxygen for 316FR, P122 and ODS steels at 550 and 650 degrees centigrade. 1. No significant dissolution attack occurred on 316FR and P122 steels at 550 degrees centigrade with 10 sup-4 wt% oxygen in LBE. They have Protective oxide layers at the surface, like in the experiments with 10 sup-6 wt% oxygen. At the high oxygen concentration (10 sup-4 wt%) thick oxide multilayers develop on 316FR instead of thin spinel layers at 10 sup-6 wt%, in both experiments LBE is enclosed in the oxide layer. With 10 sup-8 at% oxygen in LBE, however, there is no protective scale formation but strong dissolution attack (up to 120 um) occurs at the 316FR specimen.At the P122 only very little dissolution attack ($$<$$2um) is observed at the surface.

JAEA Reports

Study on Pb-Bi Corrosion of Structural and Fuel Cladding Materials for Nuclear Applications,3; Corrosion investigation of steels between 500 and 650$$^{circ}$$C during 10,000 h of exposure to stagnant liquid Pb-Bi containing 10$$^{-6}$$ wt% of oxygen

Muller, G.*; Schumacher, G.*; Weisenburger, A.*; Heinzel, A.*; Zimmermann, F.*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2003-026, 58 Pages, 2004/01

JNC-TY9400-2003-026.pdf:2.25MB

This is the third report on the compatibility of structural and fuel cladding materials that has to be investigated for a possible advanced heavy metal cooled reactor system.The first two reports considered the behavior of 316 FR, P122 and ODS steels during 800, 2,000 and 5,000 h in stagnant LBE at 500 - 650$$^{circ}$$C containing 10-6 wt% of oxygen. This report describes the results of all experiments including that of 10,000 h duration. Martensitic steels perform well at 500 and 550$$^{circ}$$C up to the maximal exposure time of 10,000 h, while austenites fail after 5,000 h. At temperatures of 600$$^{circ}$$C and 650$$^{circ}$$C all of the steels fail the earlier the higher the temperature is. Steels with Al - alloying at the surface by the GESA process withstand corrosion up to the maximal exposure time at all of the applied temperatures if it is ensured that the Al - concentration at the surface is in between 8 - 15 wt%.

Oral presentation

Fusion technology activities through the Broader Approach IFMIF-EVEDA project

Matsumoto, Hiroshi; Knaster, J.*; Heidinger, R.*; Sugimoto, Masayoshi; Ibarra, A.*; Mosnier, A.*; Heinzel, V.*; Massaut, V.*; Micciche, G.*; M$"o$slang, A.*

no journal, , 

The International Fusion Materials Irradiation Facility (IFMIF) Engineering Design and Engineering Validation Activities (EVEDA) started in 2007 under the framework of the Broader Approach (BA) Agreement between EU and Japan with the objective of developing a complete engineering design of the IFMIF together with accompanying sub projects to validate the major elements of the technologies essential for the IFMIF Plant. The validation sub projects include design and construction of the prototype deuteron beam accelerator, Lithium Loop Test Facility, and irradiation test of samples in Rigs for High Flux Test Modules. The recent achievements from these activities will be presented.

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