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Journal Articles

Corrosion resistance of Al-alloying high Cr-ODS steels in stagnant lead-bismuth

Takaya, Shigeru; Furukawa, Tomohiro; Inoue, Masaki; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Kimura, Akihiko*

Journal of Nuclear Materials, 398(1-3), p.132 - 138, 2010/03

 Times Cited Count:43 Percentile:94.69(Materials Science, Multidisciplinary)

Oxide dispersion strengthened (ODS) ferritic steels with excellent high-temperature strength are the candidates for fuel cladding tubes. But, the compatibility with lead bismuth eutectic (LBE) is one of the key issues in accelerator driven system and LBE cooled fast reactors. Addition of Al and increase in Cr may have beneficial influence on the compatibility. Addition of Al, however, causes a decrease in high-temperature strength. A significantly higher Cr concentration results in aging embrittlement. Therefore, we need to find their optimal amount to balance corrosion resistance with high-temperature strength. In this study, the cross sections of the samples after 3,000 h of exposure to LBE with 10$$^{-8}$$ wt% oxygen at 650 $$^{circ}$$C are examined in detail using scanning electron microscope and Auger electron spectroscopy. The observation shows that very thin Al oxide layer is formed continuously between multiple oxide layer/internal oxide zone and matrix, and that such Al oxide layer suppresses further growth of multiple oxide layer/internal oxide zone. The average oxide layer thickness shows a tendency to get thinner by increasing in Al content from about 2 to 4 wt%, although significant dependency on Cr content is not recognized. Furthermore, the additional corrosion test for 5,000 h is conducted. These materials show good corrosion resistance even after 5,000 h of exposure to LBE containing 10$$^{-6}$$ wt% at 650 $$^{circ}$$C. Addition of 3.5 wt% Al is very effective in improving corrosion resistance.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 4; Mechanical properties at elevated temperatures

Furukawa, Tomohiro; Otsuka, Satoshi; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*; Kimura, Akihiko*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9221_1 - 9221_7, 2009/05

As fuel cladding material for lead bismuth-cooled fast reactors and supercritical pressurized water-cooled fast reactors, our research group has been developing highly corrosion-resistant oxide dispersion strengthened ferritic steels with superior high-temperature strength. In this study, the mechanical properties of super ODS steel candidates at elevated temperature have been evaluated. Tensile tests, creep tests and low cycle fatigue tests were carried out for a total of 21 types of super ODS steel candidates which have a basic chemical composition of Fe-16Cr-4Al-0.1Ti-0.35Y$$_{2}$$O$$_{3}$$, with small variations. The testing temperatures were 700$$^{circ}$$C (for tensile, creep and low cycle fatigue tests) and 450$$^{circ}$$C (for tensile test). The major alloying parameters of the candidate materials were the compositions of Cr, Al, W and the minor elements such as Hf, Zr and Ce etc. The addition of the minor elements is considered effective in the control of the formation of the YAl complex oxides, which improves high-temperature strength. The addition of Al was very effective for the improvement of corrosion resistance. However, the addition also caused a reduction in high-temperature tensile strength. Among the efforts aimed at increasing high-temperature strength, such as the low-temperature hot-extrusion process, solution strengthening by W and the addition of minor elements, a remarkable improvement of strength was observed in ODS steel with a basic chemical composition of 2W-0.6Hf steel (SOC-14) or 2W-0.6Zr steel (SOC-16). The same behavior was also observed in creep tests, and the creep rupture times of SOC-14 and SOC-16 at 700$$^{circ}$$C - 100MPa were greater than 10,000 h. The strength was similar to that of no-Al ODS steels. No detrimental effect by the additional elements on low-cycle fatigue strength was observed in this study. These results showed that the addition of Hf/Zr to ODS-Al steels was effective in improving high-temperature strength.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 1; Introduction and alloy design

Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Lee, J. H.*; Muthukumar, N.*; Okuda, Takanari*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9220_1 - 9220_8, 2009/05

Cladding material development is essential for realization of highly efficient high burn-up operation of next generation nuclear systems, where high performance is required for the materials, that is, high strength at elevated temperature, high resistance to corrosion and high resistance to irradiation. Oxide dispersion strengthening (ODS) ferritic steels are considered to be most adequate for the cladding material because of their high strength at elevated temperature. In this work, "Super ODS steel" that has better corrosion resistance than 9Cr-ODS steel, has been developed for application to cladding of a variety of next generation nuclear systems. In the following ten papers, the recent experimental results of "Super ODS steel" R&D will be presented, indicating that many unexpected preferable features were found in the mechanical properties of nano-sized oxide dispersion high-Cr ODS ferritic steel. A series of paper begins with alloy design of "Super ODS steel". Corrosion issue requires Cr concentration more than 14wt.%, but aging embrittlement issue requires less than 16wt.%. An addition of 4wt.%Al is effective to improve corrosion resistance of 16wt.%Cr-ODS steel in supercritical water (SCW) and lead-bismuth eutectic (LBE), while it is detrimental to high-temperature strength. Additions of 2wt.%W and 0.1wt.%Ti are necessary to keep high strength at elevated temperatures. An addition of small amount of Zr or Hf results in a significant increase in creep strength at 700 $$^{circ}$$C in Al added ODS steels. Tube manufacturing was successfully done for the super ODS steel candidates. "Super ODS steel" is promising for the fuel cladding material of next generation nuclear systems, and the R&D is now ready to proceed to the next stage of empirical verification.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 2; Effect of minor alloying elements

Onuki, Somei*; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Kimura, Akihiko*; Inoue, Masaki; Kaito, Takeji; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9306_1 - 9306_5, 2009/05

For development of advanced ferritic ODS steels including high concentration of Cr and Al, the effect of minor alloying elements on fine dispersion of oxide particle was investigated. Microstructural analysis for Fe-16Cr-4Al-mY$$_{2}$$O$$_{3}$$-nZr or mHf due to TEM indicated that 0.3Zr or 0.6Hf are the optimum concentration. The mechanism of nano-sized oxide formation was also discussed.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 3; Development of high performance attrition type ball mill

Okuda, Takanari*; Fujiwara, Masayuki*; Nakai, Tatsuyoshi*; Shibata, Kenichi*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; Onuki, Somei*; Fujisawa, Toshiharu*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9229_1 - 9229_4, 2009/05

Oxygen content in ODS ferritic steel is the most important element to determine the mechanical properties. The oxygen contamination from the air is perfectly prevented by using new designed ball mill and the subsequent process control. Zr, Hf and Ti added ODS steels with three oxygen levels for the evaluation tests are fabricated.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 6; Corrosion behavior in SCPW

Lee, J. H.*; Kimura, Akihiko*; Kasada, Ryuta*; Iwata, Noriyuki*; Kishimoto, Hirotatsu*; Zhang, C. H.*; Isselin, J.*; Dou, P.*; Muthukumar, N.*; Okuda, Takanari*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9223_1 - 9223_6, 2009/05

Corrosion is a critical issue for cladding materials, especially, in sever corrosion environment as supercritical pressurized water (SCPW). In this work, the effects of alloy elements on the corrosion behavior in SCPW were investigated for a series of oxide dispersion strengthened (ODS) steels to design alloy compositions for corrosion resistant super ODS ferritic steels. Corrosion tests were carried out for the ODS steels with different concentrations of Cr and Al in SCPW at 773 K at 25 MPa with 8 ppm of dissolved oxygen. The corrosion rate of SUS430, which contained 16wt.%Cr, was much higher than 16Cr-ODS steel. This suggests that nano-sized oxide particles dispersion and very fine grains play an important role in suppression of the corrosion. The corrosion of the ODS steel was reduced by an addition of Al in 16wt.%Cr-ODS steel but not in 19Cr-ODS steel. FE-EPMA chemical analysis clearly indicated that the surface of the Al added ODS steels was covered by alumina which suppresses the corrosion in SCPW. It is considered that an adequate combination of the contents of Cr and Al is ranging (14-16)Cr and (3.5-4.5)Al.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 5; Mechanical properties and microstructure

Kasada, Ryuta*; Lee, S. G.*; Lee, J. H.*; Omura, Takamasa*; Zhang, C. H.*; Dou, P.*; Isselin, J.*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; et al.

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9072_1 - 9072_5, 2009/05

The newly-developed Al-added ODS ferritic steels with an addition of Zr or Hf, socalled super ODS candidate steels, showed good notch-impact properties in the as-received condition with keeping the excellent creep strength.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 7; Corrosion behavior and mechanism in LBE

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Kimura, Akihiko*; Inoue, Masaki; Ukai, Shigeharu*; Onuki, Somei*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9308_1 - 9308_5, 2009/05

Corrosion of structural materials is one of the serious problems when lead-bismuth eutectic alloy (LBE) is used as a coolant material in next generation nuclear systems. In this study, dissolution experiments of synthetic Fe-Cr-Al alloys and developed super ODS steel candidates into LBE under several partial pressures of oxygen were conducted. Dissolution behaviors of major components in such steels into LBE were investigated. Interfacial behavior between LBE and steels was also observed. In addition, partial potential diagrams of the Fe-Cr-Al-O system at several conditions were established as basic data. From the potential diagrams, the partial pressure range of oxygen was estimated for the stable protective oxide layer formation at the interface. At lower oxygen partial pressure than the pressure that is enough for the formation of the stable oxide layer, a rough oxide layer was formed at the interface in all samples, and the alloy elements dissolved into LBE through it. On the other hand, at the oxygen partial pressure to form stable oxide layer, a dense and very thin oxide layer was formed especially on the higher aluminum content steel, preventing the alloy dissolution into LBE. From the results, aluminum and chromium content in steel were very important for preventing the corrosion by LBE.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 8; Ion irradiation effects at elevated temperatures

Kishimoto, Hirotatsu*; Kasada, Ryuta*; Kimura, Akihiko*; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9219_1 - 9219_8, 2009/05

The Super ODS steels, having excellent high-temperature strength and highly corrosion resistant, are considered to increase the energy efficiency by higher temperature operation and extend the lifetime of next generation nuclear systems. High-temperature strength of the ODS steels strongly depends on the dispersion of oxide particles, therefore, the irradiation effect on the dispersed oxides is critical in the material development. In the present research, ion irradiation experiments were employed to investigate microstructural stability under the irradiation environment at elevated temperatures. Ion irradiation experiments were performed with 6.4 MeV Fe ions irradiated at 650 $$^{circ}$$C up to a nominal displacement damage of 60 dpa. Microstructural investigation was carried out using TEM and EDX. No significant change of grains and grain boundaries was observed by TEM investigation after the ion irradiation. Main oxide particles in the 16Cr-4Al-0.1Ti (SOC-1) ODS steel were (Y, Al) complex oxides. (Y, Ti) complex oxides were in 16Cr-0.1Ti (SOC-5) and 15.5Cr-2W-0.1Ti (SOCP-3). (Y, Zr) complex oxides were in 15.5Cr-4Al-0.6Zr (SOCP-1). No significant modification of these complex oxides was detected after the ion irradiation up to 60 dpa at 650 $$^{circ}$$C. The stable complex oxides are considered to keep highly microstructural stability of the Super ODS steels under the irradiation environments.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 9; Damage structure evolution under electron-irradiation

Onuki, Somei*; Hashimoto, Naoyuki*; Ukai, Shigeharu*; Kimura, Akihiko*; Inoue, Masaki; Kaito, Takeji; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9307_1 - 9307_4, 2009/05

The aim of this study is to survey microstructural properties and irradiation response of advanced high Cr and Al ODS steels. The effects of minor element addition and heat-treatment are also investigated. In these steels, black dots-like dislocation loops were formed around oxide particles during electron irradiation, and then the behavior depended on the type of additional elements, but the irradiation resistance was confirmed generally. The irradiation response was not sensitive as the heat-treatment, but the minor element addition (Zr and Hf) showed an intensive suppressing the loop growth. The results suggest that a large number of oxides enhanced the mutual recombination of the irradiation-induced point defects, especially at their surface.

Journal Articles

Super ODS steels R&D for fuel cladding of next generation nuclear systems, 10; Cladding tube manufacturing and summary

Ukai, Shigeharu*; Onuki, Somei*; Hayashi, Shigenari*; Kaito, Takeji; Inoue, Masaki; Kimura, Akihiko*; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9232_1 - 9232_7, 2009/05

The super ODS cladding were manufactured into 8.5 mm outer diameter and 0.5 mm thickness by using pilger.mill rolling and intermediate heat treatment. The final heat treatment was successfully conducted at 1150 $$^{circ}$$C for 1h to make perfectly recrystallized structure. The manufactured cladding exhibits a secondary recrystallized grains with {110} $$<$$100$$>$$ Goss orientation. As a summary of super ODS steels R&D, the ODS steel, 16Cr-4Al-2W-0.1Ti added with small amount of Hf or Zr, is the prime candidate of "Super ODS steels" which has high strength at elevated temperatures, high resistance to corrosion for SCW and LBE, and high resistance to irradiation.

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.

JAEA Reports

Solubility of Major Elements of Steel into Lead-Bismuth Alloy

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2005-007, 27 Pages, 2005/03

JNC-TY9400-2005-007.pdf:1.05MB

Lead-Bismuth eutectic alloy (LBE) has been considered as a prospective coolant for a fast-breeder reactor, but a corrosion of cooling pipe is anticipated when it is used. In the previous study, the solubility of major metallic elements, such as Fe, Cr and Ni, into LBE was measured under extra low oxygen potential. The interactive effect of those elements on the solubility was also examined. However, it is thought that an oxide layer is formed on the surface of the cooling pipe and influences on the corrosion of the pipe. In this year, the measurement of the solubility of major elements of steel into lead-bismuth alloy through Fe-Cr complex oxide was started. In addition, the thermodynamics of the Pb-Bi-O system at 800 K was investigated and the phase diagram of the system was determined. From the above results, the stability diagram of the Pb-Bi-O system was established.

JAEA Reports

Solubility of metallic elements in LBE under extra low oxygen potential (JFY2003 joint research report)

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2004-015, 55 Pages, 2004/03

JNC-TY9400-2004-015.pdf:1.45MB

In this study, solubility of major metallic elements in LBE was measured under extra low oxygen potential.

JAEA Reports

Solubility of metallic elements in LBE under extra low oxygen potential; JFY2001 joint research report

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Furukawa, Tomohiro; Aoto, Kazumi

JNC TY9400 2003-002, 22 Pages, 2003/03

JNC-TY9400-2003-002.pdf:0.57MB

Lead-Bismuth eutectic alloy(LBE) has been considered as a prospective coolant for a fast-breeder reactor. However a corrosion of the structual material is anticipated when it is used at the similar temperature as sodium coolant. In this study, solubility of major metallic elements in LBE is measured under extra low oxygen potential. The interactive effect of those elements on the solubility is also to be examined. Based on the results of JFY2001, measurements of the solubility of iron in LBE at 1273 K were conducted where the partial pressure of oxygen was controlled by using two equilibrium methods: the oxide equilibrium method and the gas equilibrium method. Analytical condition of oxygen content in LBE was also determined. Several solubility data were obtained. However it was not yet enough to do thermodynamic consideration. From the above results, following subjects were extracted for JFY2003 study. (1)To accumulate the solubility data of iron and oxygen in LBE and to consider them thermodynamically. (2)To obtain the solubility data of the other composition element of stainless steels and to evaluate mutual influence of them.

Oral presentation

Solubility of (Fe,Cr)$$_{3}$$O$$_{4}$$ into Molten Pb-Bi eutectic

Sano, Hiroyuki*; Fujisawa, Toshiharu*; Sagawa, Yosuke*; Furukawa, Tomohiro; Aoto, Kazumi

no journal, , 

no abstracts in English

Oral presentation

Out of pile FCCI susceptibility test of super ODS ferritic steel

Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Okuda, Takanari*; Kimura, Akihiko*; Ukai, Shigeharu*; Onuki, Somei*; Fujisawa, Toshiharu*; Abe, Fujio*

no journal, , 

Corrosion-resistant super ODS ferritic steels are being developed for high-burn up fuel cladding material for LBE-cooled fast reactor and SCW-cooled fast reactor. One of the critical tasks is to ensure the resistance to fuel-cladding chemical interaction(FCCI) of the cladding material. JAEA reseraches showed that the volatile elements Cs, Te and I play an important role in FCCI behavior, and that out of pile corrosion test using the CsOH/CsI mixture can provide the similar corrosion morphology to the actual inner face corrosion of spent fuel pin. This study elaluated the effects of diffenrence in Cr and Al concentrations brought on corrosion resistance to Cs and I. The corrosion resistance was improved by either increase of Cr concentration or removal of Al. The increase of Cr concentration is the key to improvement of FCCI resistance and can produce a satisfactory FCCI resistance surpassing SUS316, even if Al is added to the steel.

Oral presentation

Super ODS steel R&D towards highly efficient nuclear systems, 2; Mechanical properties and nano-meso structure control

Furukawa, Tomohiro; Otsuka, Satoshi; Inoue, Masaki; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*; Kimura, Akihiko*

no journal, , 

High temperature tensile, creep, high temperature low cycle fatigue and impact tests have been performed for 21 kinds of Super ODS Candidate (SOC) steel. The mechanical strength of the SOC steels which added aluminum effective in corrosion resistant in LBE/SCW was lower than that of the other steel which did not add aluminum. However, it turned out that low temperature extrusion, addition of tungsten into the steel, addition of the active metal elements such as hafnium and zirconium are effective in an improvement of high temperature mechanical strength.

Oral presentation

Super ODS steel R&D towards highly efficient nuclear systems, 4; Corrosion behavior in LBE

Takaya, Shigeru; Furukawa, Tomohiro; Aoto, Kazumi; Inoue, Masaki; Fujisawa, Toshiharu*; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Kimura, Akihiko*

no journal, , 

We report the results of long-term corrosion test for evaluation of corrosion resistance of super ODS steel candidates in LBE.

Oral presentation

Super ODS steel R&D towards highly efficient nuclear systems, 6; Piping and summary

Kaito, Takeji; Inoue, Masaki; Ukai, Shigeharu*; Kimura, Akihiko*; Okuda, Takanari*; Abe, Fujio*; Onuki, Somei*; Fujisawa, Toshiharu*

no journal, , 

The Four path tube manufacturing of the corrosion resistant super ODS ferritic steel was carried out using pilger mill cold working and intermediate softening heat treatment. High-temperature final heat treatment successfully produced isotropical recrystallized microstructure. It was confirmed that target precise dimension of 8.5mm in outer diameter $$times$$ 0.5mm in wall thickness can be achieved by the tube manufucturing process developed by JAEA. Based on the reviewing of R&D results on super ODS steel, the prospect of super ODS steel development was discussed.

29 (Records 1-20 displayed on this page)