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Journal Articles

Manufacturability and strength assessment of modified 9Cr-1Mo steel products for Japan sodium cooled fast reactor components, 2; Long thin-walled tubes with small diameter

Wakai, Takashi; Onizawa, Takashi; Obara, Satoshi; Nakashima, Takashi*; Yokoyama, Tetsuo*; Iseda, Atsuro*; Ogumo, Shinya*; Futagami, Satoshi*

Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 52(2), p.171 - 181, 2011/07

To enhance the economic competitiveness, high-Cr ferritic steels are adopted as the structural materials for JSFR, because the steels have both excellent high temperature strength and thermal properties. Among the high-Cr ferritic steels, modified 9Cr-1Mo steel (compatible to ASTM A213 T91) is a candidate of the structural material for the demonstration facility of JSFR, because the steel has superior microstructure stability at elevated temperature for long time. As for the steam generator tubes, to enhance the safety, straight double-walled tubes will be employed. In this study, the following technical issues were investigated; (1) Industrial manufacturability of thin-walled small bore tubes made of modified 9Cr-1Mo steel, (2) performance demonstration of the thin-walled small bore tubes, (3) industrial manufacturability of double-walled tubes using the thin-walled small bore tubes and (4) extraction of technical problems to manufacture the double-walled tubes for the JSFR steam generator. As a result, thin-walled small bore tubes made of modified 9Cr-1Mo steel were successfully manufactured by using the existing industrial facilities, up to 17m in length. The mechanical properties of the tubes satisfy the requirements from plant designing, as well as those from the material strength standards. Further, double walled tubes were also manufactured, up to 15m in length. The surface conditions of the tubes and the contact pressure between inner and outer tubes should be optimized.

Journal Articles

The Creep rupture strength evaluation in welded joint of Mod.9Cr-1Mo steel

Wakai, Takashi; Nagae, Yuji; Takaya, Shigeru; Obara, Satoshi; Date, Shingo*; Yamamoto, Kenji*; Kikuchi, Koichi*; Sato, Kenichiro*

Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 52(2), p.147 - 159, 2011/07

By employing high-Cr ferritic steels to the structural materials for JSFR, a compact plant designing can be achieved. It contributes to reduce the construction cost and to enhance the freedom of designing. Among the high-Cr ferritic steels, modified 9Cr-1Mo steel (compatible to ASTM A335 P91) is a candidate of the structural material for the demonstration facility of JSFR, because the steel has superior microstructure stability at elevated temperature for long time. However, remarkable creep strength degradation has been observed in the welded joint of high-Cr ferritic steels, especially in long-term region. It is known as "Type-IV damage". In the elevated temperature designing for the fast reactors, such creep strength degradation must be taken into account properly. Therefore, the creep strength assessment procedure and the allowable stress for the welded joints made of modified 9Cr-1Mo steel have been proposed. In this study, (1) a series of creep rupture tests to verify the validity of the creep strength assessment procedure was performed. (2) Applicability of the creep strength assessment procedure to the creep fatigue strength assessment of the welded joints made of modified 9Cr-1Mo steel was investigated. (3) Metallurgical examinations of creep ruptured specimens were carried out to confirm the contribution of "Type-IV damage". As a result, it was demonstrated that the creep strength assessment procedure was validated using the long-term creep rupture test results less than 30,000h and that the creep strength assessment procedure was applicable to the creep-fatigue strength assessment based on some uniaxial creep-fatigue test results.

Journal Articles

Influence of FBR plant service and repair welding on microstructure and residual stress of austenitic stainless steel weld joint

Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru; Martin, L.*

Kensa Gijutsu, 16(3), p.24 - 30, 2011/03

For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545$$^{circ}$$C in the French proto-type fast reactor Phenix (as secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.

Journal Articles

A Crossed photon-atom beam method for absolute measurement of total photoionization cross sections on isolated metal atoms; Measurements on Ba and Eu atoms

Obara, Satoshi*; Kobayashi, Ryoei*; Yagi, Shuichi*; Toyama, Yuko*; Kutluk, G.*; Osawa, Tetsutaro*; Ogura, Koichi; Shibata, Takemasa; Azuma, Yoshiro*; Nagata, Tetsuo*

Nuclear Instruments and Methods in Physics Research B, 269(3), p.263 - 271, 2011/02

 Times Cited Count:0 Percentile:0.01(Instruments & Instrumentation)

A crossed photon-atom beam apparatus has been constructed for absolute measurement of total photoionization cross sections of isolated and neutral metallic atoms. Using this apparatus, measurements on Ba and Eu atoms have been made at their 4d giant resonance regions 110-140 eV and 140-180 eV, respectively. The target atom density was determined using the deposition rate on a quartz crystal sensor and the average velocity of the atoms obtained by a time-of-flight method combined with a pulsed electron gun. The number of photons was determined with use of a double ion chamber. The comparison of the measured cross-section values with previous experimental and theoretical results is reasonable, indicating that the crossed photon-atom beam method is fairly promising technique.

Journal Articles

Creep strength evaluation of welded joint made of modified 9Cr-1Mo steel for Japanese Sodium Cooled Fast Reactor (JSFR)

Wakai, Takashi; Nagae, Yuji; Onizawa, Takashi; Obara, Satoshi; Xu, Y.*; Otani, Tomomi*; Date, Shingo*; Asayama, Tai

Proceedings of 2010 ASME Pressure Vessels and Piping Conference (PVP 2010) (CD-ROM), 9 Pages, 2010/07

Journal Articles

Influence of FBR plant service and repair welding on microstructure and residual stress of austenitic stainless steel weld joint

Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru*; Martin, L.*

Hozengaku, 9(1), p.32 - 38, 2010/04

For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545$$^{circ}$$C in the French proto-type fast reactor Phenix (secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.

Journal Articles

Influence of heat treatment on long term creep properties of 9Cr-W-Mo-V-Nb steel

Obara, Satoshi; Wakai, Takashi; Asayama, Tai; Yamada, Yoshiyuki*; Nakazawa, Takanori*; Yamazaki, Masayoshi*; Hongo, Hiromichi*

Tetsu To Hagane, 96(4), p.172 - 181, 2010/04

This paper describes the effect of heat treatment on mechanical properties of 9Cr-W-Mo-V-Nb steel as a part of development of high Cr steel for fast breeder reactor (FBR). The effects of normalizing temperature and tempering temperature and time on creep properties were investigated from the viewpoint of microstructures. Creep strength increased with increase in normalizing temperature from 950$$^{circ}$$C to 1200$$^{circ}$$C. The microstructural factors that improve creep strength are increased amount of MX which precipitated during tempering process and increased dislocation density. In addition, coarsening of prior austenite grain size also contribute to increasing creep strength. Although creep rupture time of 780$$^{circ}$$C-1h tempering specimen was greater than that of 700$$^{circ}$$C-1h and 750$$^{circ}$$C-1h tempering specimen in a long-term region exceeding 20000h at 650$$^{circ}$$C, creep ductility and reduction of area of the former remarkably degraded compared to those of the latter.

Journal Articles

Influence of heat treatments on high temperature tensile properties and toughness of 9Cr-W-Mo-V-Nb steel

Obara, Satoshi; Wakai, Takashi; Asayama, Tai; Yamada, Yoshiyuki*; Nakazawa, Takanori*

Tetsu To Hagane, 95(5), p.417 - 425, 2009/05

 Times Cited Count:6 Percentile:42.8(Metallurgy & Metallurgical Engineering)

This paper describes the effect of heat treatment on mechanical properties of high chromium (Cr) ferritic steel 9Cr-W-Mo-V-Nb steel as a part of development of high Cr steel for fast breeder reactor (FBR). The effects of normalizing temperature and tempering temperature and time on high temperature tensile properties and Charpy impact properties were investigated from the viewpoint of microstructures.

Journal Articles

Influence of normalising temperature on MX precipitation behaviour in high-chromium steel

Obara, Satoshi; Onizawa, Takashi; Wakai, Takashi; Asayama, Tai

Proceedings of Workshop on Structural Materials for Innovative Nuclear Systems (SMINS), p.115 - 126, 2008/07

This study investigates the influence of normalising temperature on metal carbonitride (MX) precipitation behavior and mechanical property of high chromium (Cr) ferritic steel for sodium cooled fast reactor (SFR) structures. Generally, mechanical strength in high Cr steel is increased addition of vanadium (V) and niobium (Nb), because V and Nb play a role of precipitation strengthening elements as metal carbonitride (MX) particles. However, precipitation behavior depends not only on the amount of V and Nb but also on the heat treatment conditions. In this presentation, in order to investigate the optimum heat treatment conditions for FBR structural materials, several kinds of high Cr steels with different heat treatments were produced and a series of mechanical tests and metallurgical examinations were conducted.

JAEA Reports

Study on assembly techniques and procedures for ITER tokamak device

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi*; Ue, Koichi*; Shimizu, Katsusuke*; Onozuka, Masanori*

JAEA-Technology 2006-034, 85 Pages, 2006/06

JAEA-Technology-2006-034.pdf:9.18MB

The International Thermonuclear Experimental Reactor (ITER) tokamak is composed of many kinds of components. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of these components are required to be a high accuracy of $$pm$$3mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements. Based on the above background, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology concept based on the available knowledge and experiences of the installation of the large components, in order to improve the IT (International Team) design toward the more realistic one. As a result, we show the realistic tokamak assembly procedures and techniques to be able to assemble the large and heavy ITER tokamak with high accuracy. (1)Assembly and alignment of the toroidal field coil with high accuracy. (2)Simplification of the assembly procedures, and the jigs/tools and procedures to reduce the misalignment. (3)Assembly procedures and techniques for the vacuum vessel to reduce the weld distortion. (4)Supporting procedures and techniques of the vacuum vessel sector to prevent the toridal field coil from the deformation due to the dead weight of the vacuum vessel sector. (5)Datum points and lines for the required positions and assembly tolerances during tokamak assembly.

JAEA Reports

Continuous running test of radiation resistance motor driving equipment under high gamma ray irradiation

Obara, Kenjiro; Kakudate, Satoshi; Yagi, Toshiaki; Morishita, Norio; Shibanuma, Kiyoshi

JAEA-Technology 2006-023, 38 Pages, 2006/03

JAEA-Technology-2006-023.pdf:5.21MB

The components in the vacuum vessel of ITER (International Thermonuclear Experimental Reactor), e.g. blanket and divertor, are replaced using the dedicated remote handling systems. The environment conditions inside the vacuum vessel during the operation are temperature of 50$$^{circ}$$C, gamma ray radiation and air or inert gas atmosphere at 1 atm. ; therefore multiple elements are required as durability of the remote handling systems. In addition, the remote handling system it is desired to be able to operate over a long time. The radiation resistance motor driving equipment, which comprises parts with different radiation resistance levels, was designed simulating mechanisms of the ITER remote handling systems. The equipment being the servomotor, turns the weight (dummy load) of 8 kgf and controls, and continuous running test under high gamma ray irradiation was started from March, 2000. Irradiation conditions on the test were the dose rate of 3.6 kGy/h, the target accumulation dose of 30 MGy at the minimum. The irradiation test was performed two stages which was divided by overhaul of the equipment. The achieved accumulation dose and running time in these stages were approximately 47.6 MGy/13,200 hours and 23.9 MGy/6,640 hours, respectively. As a result, it has been confirmed that sufficient radiation resistance of the equipment, which is required from the latest dose rate of 0.5 kGy/h inside the vacuum vessel was achieved. In this report, we describe design conditions of the equipment and the results of the 1st and 2nd irradiation tests and the overhaul after the 1st irradiation test.

JAEA Reports

Operation and management data for the JRR-3 CNS Facility

Suzuki, Masatoshi; Hazawa, Tomoya; Ishizaki, Yoichi*; Obara, Michio*; Inada, Katsutoshi*; Yonekawa, Mitsunori*; Wakita, Hiroshi*

JAERI-Tech 2004-060, 153 Pages, 2004/09

JAERI-Tech-2004-060.pdf:9.78MB

The cold neutron source (CNS) at JRR-3 was constructed for the purpose of improving the utilization performance of the reactor along the lines of upgrade program. There are two methods for extracting cold neutrons from a reactor, one is to filter a small fraction of cold neutrons in the Maxwellian spectrum of reactor neutrons while the other is to increase the fraction of cold neutron by inserting a cryogenic moderator. The latter is adopted as CNS facilities at almost all cases, and liquid hydrogen and its cooling system are equipped with a cold neutron source at many neutron facilities just like JRR-3. Cold neutrons generated in a cold source are extracted through a neutron guide tube, and are utilized for the purpose of neutron beam experiments such as neutron scattering which study the structures of atoms and molecules in the materials and life science fields. This report summarizes the operation data and the main technical issues which were recorded in the whole operation period from commencement date in 1989 to March 2004.

JAEA Reports

Development of fabrication technology for ITER vacuum vessel

Nakahira, Masataka; Shibanuma, Kiyoshi; Kajiura, Soji*; Shibui, Masanao*; Koizumi, Koichi; Takeda, Nobukazu; Kakudate, Satoshi; Taguchi, Ko*; Oka, Kiyoshi; Obara, Kenjiro; et al.

JAERI-Tech 2002-029, 27 Pages, 2002/03

JAERI-Tech-2002-029.pdf:2.04MB

The ITER vacuum vessel (VV) R&D has progressed with the international collaborative efforts by the Japan, Russia and US Parties during the Engineering Design Activities (EDA). Fabrication and testing of a full-scale VV sector model and a port extension have yielded critical information on the fabrication and assembly technologies of the vacuum vessel, magnitude of welding distortions, dimensional accuracy and achievable tolerances during sector fabrication and field assembly. In particular, the dimensional tolerances of $$pm$$3 mm for VV sector fabrication and $$pm$$10 mm for VV sector field assembly have been achieved and satisfied the requirements of $$pm$$5 mm and $$pm$$20 mm, respectively. Also, the basic feasibility of the remote welding robot has been demonstrated. This report presents detailed fabrication and assembly technologies such as welding technology applicable to the thick wall without large distortion, field joint welding technology between sectors and remote welding technology through the VV R&D project.

Journal Articles

Mechanical characteristics and position control of vehicle/manipulator for ITER blanket remote maintenance

Kakudate, Satoshi; Oka, Kiyoshi; Yoshimi, Takashi*; Taguchi, Ko*; Nakahira, Masataka; Takeda, Nobukazu; Shibanuma, Kiyoshi; Obara, Kenjiro; Tada, Eisuke; Matsumoto, Yasuhiro*; et al.

Fusion Engineering and Design, 51-52(1-4), p.993 - 999, 2000/11

 Times Cited Count:9 Percentile:55.69(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

Obara, Kenjiro; Kakudate, Satoshi; *; Shibanuma, Kiyoshi; Tada, Eisuke

JAERI-Tech 99-009, 83 Pages, 1999/02

JAERI-Tech-99-009.pdf:5.99MB

no abstracts in English

JAEA Reports

High $$gamma$$-rays irradiation tests of critical components for ITER (International Thermonulear Experimental Reactor) in-vessel remote handling system

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; *; *; Koizumi, Koichi; Shibanuma, Kiyoshi; Yagi, Toshiaki; Morita, Yosuke; ; et al.

JAERI-Tech 99-003, 312 Pages, 1999/02

[This article is unavailable to download the full text due to various reasons.]

Journal Articles

Development of 15-m-long radiation hard periscope for ITER in-vessel viewing

Obara, Kenjiro; *; Kakudate, Satoshi; Oka, Kiyoshi; Nakahira, Masataka; Morita, Yosuke; *; *; Takeda, Nobukazu; Takahashi, Hiroyuki*; et al.

Fusion Engineering and Design, 42, p.501 - 509, 1998/00

 Times Cited Count:3 Percentile:32.94(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Development of radiation hardness components for ITER remote maintenance

Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; *; Yagi, Toshiaki; Morita, Yosuke

J. Robot. Mechatron., 10(2), p.121 - 132, 1998/00

no abstracts in English

Journal Articles

Development of ITER in-vessel viewing and metrology systems

Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; *

J. Robot. Mechatron., 10(2), p.96 - 103, 1998/00

no abstracts in English

Journal Articles

Development of bore tools for blanket cooling pipe connection in ITER

*; Oka, Kiyoshi; Kakudate, Satoshi; Obara, Kenjiro; *; Tada, Eisuke; A.Tesini*; Shibanuma, Kiyoshi; R.Haange*

Proc. of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 2, p.921 - 924, 1998/00

no abstracts in English

50 (Records 1-20 displayed on this page)