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Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Sekio, Yoshihiro; Yamagata, Ichiro; Yamashita, Shinichiro; Inoue, Masaki; Maeda, Koji
Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 10 Pages, 2014/10
In the Fukushima Dai-ichi Nuclear Power Plant accident, seawater was temporarily injected into the spent fuel pools since water cooling and feeding functions were lost. For fuel assemblies which experienced seawater immersion, surface corrosion due to seawater constituents and the resultant degradation of mechanical property are of concern. Therefore, in order to assess the integrity of fuel assemblies (especially cladding tubes), the effects of seawater immersion on corrosion behavior and mechanical properties for as-recieved and irradiated Zircaloy-2 cladding tubes were investigated in the present study. As a result, no obvious surface corrosion and no significant degradation in the tensile strength property were observed after both artificial and natural seawater immersion tests for both steels. This suggests that the effects of seawater immersions on corrosion behavior and mechanical property (especially tensile property) for Zircaloy-2 cladding tubes are probably negligible.
Inoue, Masaki; Asaka, Takeo
JAEA-Review 2014-020, 46 Pages, 2014/06
JAEA-Review-2014-020.pdf:5.81MB
Integrity of fuel assemblies (FAs) stored in the spent fuel pools (SFPs) of Fukushima Dai-ichi Nuclear Power Station (units 1 through 4) is one of the most important issues to transport the FAs to the common pool for long term storage. The SFPs had lost their functions of decay heat removal and water supply due to the station blackout. Since fresh and sea waters were injected into and concrete fragments by hydrogen explosions fell into the SFPs, the FAs have been exposed to much more corrosive environments than usual ones. In this report, many events during the accidents were investigated from a view point of temperature and chemical constituents in the SFPs in order to evaluate integrity for fuel assemblies during long term storage in the common pool by means of corrosion tests.
Inoue, Masaki; Kaito, Takeji
JAEA-Research 2013-041, 69 Pages, 2014/01
JAEA-Research-2013-041.pdf:4.61MB
Long term performance of radial shielding subassemblies with zirconium hydride, which is one of the key technologies to reduce reactor vessel radius, was evaluated for the demonstration fast breeder reactor core. Hydrogen permeation through cladding tube wall and release into primary coolant is essential to design cold traps and shielding performance. Also, higher thermal neutron fluence produces larger helium in cladding tube steels, and may degrade mechanical properties and dimensional stability. A new model was established to quantitatively calculate hydrogen release and helium production under steep gradient of neutron and 
ray fluxes in outer core region. Austenitic stainless steel (PNC316) and ferritic/martensitic steel (PNC-FMS) will not be capable for 60 years because of large helium production and high permeability, respectively. In contrast, dual wall tube combining PNC-FMS with surface oxidized Fe-18Cr-2Al alloy will be applicable for 60 years in case that manufacturing process is successfully developed.
Yamagata, Ichiro; Hayashi, Takehiro; Mashiko, Shinichi*; Sasaki, Shinji; Inoue, Masaki; Yamashita, Shinichiro; Maeda, Koji
JAEA-Testing 2013-004, 23 Pages, 2013/11
JAEA-Testing-2013-004.pdf:8.59MB
In the accident of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Co. accompanying the Great East Japan Earthquake, fuel assemblies kept in the spent fuel pool of reactor units 1-4, were exposed to the inconceivable environment such as falling and mixing of rubble, especially seawater were injected into unit 2-4. In order to evaluate the integrity of the fuel assemblies in spent fuel pools, and in the long-term storage after transported to the common storage pool, the immersion tests were performed using zircaloy-2 fuel cladding tubes irradiated in the advanced thermal reactor Fugen. The immersion liquid was prepared with doubling dilution of artificial seawater, which temperature was 80 
C and immersion time was about 336 hours, as assuming the situation of the pool. The results indicated zircaloy-2 cladding tubes had no significant corrosion and no influence on mechanical property by immersion tests with artificial seawater conditions of this work.
Inoue, Masaki; Suto, Mitsuo; Koyama, Shinichi; Otsuka, Satoshi; Kaito, Takeji
JAEA-Research 2013-009, 78 Pages, 2013/10
JAEA-Research-2013-009.pdf:3.75MB
In order to exammine the applicability for advanced aqueous reprocessing system, the martensitic oxide dispersion strengthened ferritic steel (9Cr-ODS steel), which is the primary candidate material for high burnup fuel pin cladding tube in fast reactor cycle, was evaluated for the amount of corrosion products in the dissolution process. The quantity of corrosion products was calculated to investigate the influence of both various chemical processes and waste glass (vitrified high level radioactive wastes) by use of the results of a maximum cladding temperature fuel subassembly and the sum of all fuel subassemblies, respectively. The experimental results of immersion tests in flowing liquid sodium loops and fuel pin irradiation tests in fast reactors were reviewed to consider the effect of outer and inner corrosions in high burnup fuel pins on corrosion products. This work revealed that the sum of corrosion products depends largely on the mass transfer behavior in flowing liquid sodium.
Otsuka, Satoshi; Kaito, Takeji; Ukai, Shigeharu*; Inoue, Masaki; Okuda, Takanari*; Kimura, Akihiko*
Journal of Nuclear Materials, 441(1-3), p.286 - 292, 2013/10
Times Cited Count：4 Percentile：32.48(Materials Science, Multidisciplinary)The Al addition to ODS ferritic steels considerably improves the compatibility between U-Zr fuel and the ODS steels. The threshold temperature for reaction layer formation is roughly 50K higher in the Al-containing ODS ferritic steels than in those same steels without Al addition. The activity calculation results obtained using general thermodynamic data indicate the possibility that stabilization of the intact alpha-Zr layer by Al addition is the main mechanism for the compatibility improvement by Al addition.
Kaito, Takeji; Yano, Yasuhide; Otsuka, Satoshi; Inoue, Masaki; Tanaka, Kenya; Fedoseev, A. E.*; Povstyanko, A. V.*; Novoselov, A.*
Journal of Nuclear Science and Technology, 50(4), p.387 - 399, 2013/04
Times Cited Count：14 Percentile：74.42(Nuclear Science & Technology)In order to confirm the irradiation behavior of ODS steels and thus judge their applicability to fuel claddings, fuel pin irradiation tests using 9Cr and 12Cr-ODS claddings developed by JAEA were conducted to burnup of 11.9 at% and neutron dose of 51 dpa in the BOR-60. Superior properties of the ODS claddings concerning FCCI, dimensional stability under irradiation and so on were confirmed indicating good application prospects for high burnup fuel. On the other hand, peculiar irradiation behaviors, fuel pin failure and the microstructure change containing coarse and irregular precipitates, occurred in a part of the fuel pin with 9Cr-ODS cladding. This paper describes evaluation of the obtained irradiation data and the investigation results into the cause of the peculiar irradiation behaviors.
Inoue, Masaki; Ikeuchi, Hirotomo; Takeuchi, Masayuki; Koyama, Shinichi; Suto, Mitsuo
JAEA-Research 2011-057, 100 Pages, 2012/03
JAEA-Research-2011-057.pdf:3.23MB
Corrosion resistance of fuel pin cladding tube materials is one of the most important properties to design aqueous reprocessing process. The martensitic oxide dispersion strengthened ferritic steel, names as "9Cr-ODS" steel, is the primary candidate of high burnup fuel pin cladding tube for fast reactor cycle. Because 9Cr-ODS steel contains lower chromium than stainless steels, oxidizing species in nitric acid medium needs to reduce its corrosion rate. In spent fuel dissolvers, although both nitric acid and metallic ions concentrations change, corrosion potential of 9Cr-ODS steel tends to increase gradually and stabilize protective passive layer effectively.
Yano, Yasuhide; Ogawa, Ryuichiro; Yamashita, Shinichiro; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Tanaka, Kenya
Journal of Nuclear Materials, 419(1-3), p.305 - 309, 2011/12
Times Cited Count：20 Percentile：80.18(Materials Science, Multidisciplinary)The effects of fast neutron irradiation on ring tensile properties of oxide dispersion strengthened (ODS) steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the material irradiation rig at temperatures between 693 and 1108 K to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures. The experimental results showed that there was no significant change in tensile strengths after neutron irradiation below 923 K, but the tensile strengths at neutron irradiation above 1023 K up to 33 dpa were decreased by about 20%. On the other hand, uniform elongation after irradiation was more than 2% at all irradiation conditions. The ring tensile properties of these ODS claddings remained excellent within these irradiation conditions compared with conventional 11Cr ferritic/martensitic steel (PNC-FMS) claddings.
Nogiwa, Kimihiro; Nishimura, Akihiko; Yokoyama, Atsushi; Otsuka, Satoshi; Kaito, Takeji; Inoue, Masaki; Okubo, Tadakatsu*; Hono, Kazuhiro*
Journal of Nuclear Materials, 417(1-3), p.201 - 204, 2011/10
Times Cited Count：8 Percentile：53.37(Materials Science, Multidisciplinary)Du se 9Cr-ODS (oxide dispersion-strengthened) steel consisting of residual-
ferrite and prime martensite has excellent high-temperature strength. This study describes the microstructure of dual-phase 9Cr-ODS steels characterized by atom-probe tomography in order to compare oxide-particle dispersion states in each phase. This revealed that nano-size oxide particles were of the same chemical composition and that their mean size was about 3 nm in each phase. On the other hand, the number density in the residual- phase was about four times higher than that of the prime phase. These results indicate that the dense distribution of the oxide particles in the residual- phase contribute to the excellent high-temperature strength of 9Cr-ODS steel.
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：59 Percentile：96.05(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
wt% oxygen at 650 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
wt% at 650 C. Addition of 3.5 wt% Al is very effective in improving corrosion resistance.
Onuma, Masato*; Suzuki, Junichi; Otsuka, Satoshi; Kim, S.-W.; Kaito, Takeji; Inoue, Masaki; Kitazawa, Hideaki*
Acta Materialia, 57(18), p.5571 - 5581, 2009/10
Times Cited Count：117 Percentile：97.23(Materials Science, Multidisciplinary)
-ferrite proportion on creep strength of 9Cr-ODS steelOtsuka, Satoshi; Kaito, Takeji; Kim, S.-W.; Inoue, Masaki; Asayama, Tai; Onuma, Masato*; Suzuki, Junichi
Materials Transactions, 50(7), p.1778 - 1784, 2009/06
Times Cited Count：11 Percentile：53.43(Materials Science, Multidisciplinary)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
O
, with small variations. The testing temperatures were 700C (for tensile, creep and low cycle fatigue tests) and 450C (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 700C - 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.
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 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.
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-mYO-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.
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.
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.
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.
