Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 23

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

JAEA Reports

Evaluation of irradiation behavior on oxide dispersion strengthened (ODS) steel claddings irradiated in Joyo/CMIR-6

Yano, Yasuhide; Otsuka, Satoshi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Sekine, Manabu; Endo, Toshiaki; Yamagata, Ichiro; Sekio, Yoshihiro; Tanno, Takashi; Uwaba, Tomoyuki; et al.

JAEA-Research 2013-030, 57 Pages, 2013/11

JAEA-Research-2013-030.pdf:48.2MB

It is necessary to develop the fast reactor core materials, which can achieve high-burnup operation improving safety and economical performance. Ferritic steels are expected to be good candidate core materials to achieve this objective because of their excellent void swelling resistance. Therefore, oxide dispersion strengthened (ODS) ferritic steel and 11Cr-ferritic/martensitic steel (PNC-FMS) have been respectively developed for cladding and wrapper tube materials in Japan Atomic Energy Agency. In this study, the effects of fast neutron irradiation on mechanical properties and microstructure of 9Cr-and 12Cr-ODS steel claddings for fast reactor were investigated. Specimens were irradiated in the experimental fast reactor Joyo using the CMIR-6 at temperatures between 420 and 835$$^{circ}$$C to fast neutron doses ranging from 16 to 33 dpa. The post-irradiation ring tensile tests were carried out at irradiation temperatures.

Journal Articles

Mechanical properties of friction stir welded 11Cr-ferritic/martensitic steel

Yano, Yasuhide; Sato, Yutaka*; Sekio, Yoshihiro; Otsuka, Satoshi; Kaito, Takeji; Ogawa, Ryuichiro; Kokawa, Hiroyuki*

Journal of Nuclear Materials, 442(1-3), p.S524 - S528, 2013/09

 Times Cited Count:12 Percentile:73.42(Materials Science, Multidisciplinary)

Friction stir welding was applied to the wrapper tube materials, 11Cr-ferritic/martensitic steel, intended for fast reactors and defect-free welds were successfully produced. Then, the mechanical and microstructural properties of the friction stir welded steel were investigated. The hardness values of the stir zone were about 550 Hv, and they had hardly any dependence on the rotational speed, although they were much higher than that of the base material. However, tensile strengths and elongations of the stir zones were better at 298 K, compared to those of the base material. These excellent tensile properties were attributable to the fine grain formation during friction stir welding. A part of this study is the result of "Friction stir welding of the wrapper tube materials for Na fast reactors" carried out under the Strategic Promotion Program for Basic Nuclear Research by the Ministry of Education, Culture, Sports, Science and Technology of Japan.

Journal Articles

Investigation of the cause of peculiar irradiation behavior of 9Cr-ODS steel in BOR-60 irradiation tests

Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

Journal of Nuclear Science and Technology, 50(5), p.470 - 480, 2013/05

 Times Cited Count:3 Percentile:29.69(Nuclear Science & Technology)

Four experimental fuel assemblies (EFAs) containing 9Cr-ODS steel cladding fuel pins were previously irradiated in the BOR-60. One of the EFAs achieved the best data, a peak burn-up of 11.9at% and a neutron dose of 51 dpa, without any microstructure instability or any fuel pin rupture. On the other hand, in another EFA (peak burn-up, 10.5at%; peak neutron dose, 44 dpa), peculiar irradiation behaviors such as microstructure instability and fuel pin rupture occurred. The combined effects of matrix Cr heterogeneity (presence of metallic Cr inclusions) and high-temperature irradiation were concluded to be the main cause of the peculiar microstructure change of 9Cr-ODS steel cladding tubes in the BOR-60 irradiation tests. They contributed to the fuel pin rupture.

Journal Articles

Irradiation performance of oxide dispersion strengthened (ODS) ferritic steel claddings for fast reactor fuels

Kaito, Takeji; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Yamashita, Shinichiro; Ogawa, Ryuichiro; Tanaka, Kenya

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 11 Pages, 2013/03

The oxide dispersion strengthened (ODS) ferritic steel claddings developed by Japan Atomic Energy Agency were irradiated in Joyo and BOR-60 in order to confirm their irradiation performance and thus judge their applicability to high burnup and high temperature fast reactor fuels. In Joyo, material irradiation tests up to 33 dpa were carried out at in the temperature range of 693 - 1108 K. The irradiation data were obtained concerning mainly mechanical properties and of microstructure stability. In BOR-60, fuel pin irradiation tests were conducted up to burnup of 11.9 at% and neutron dose of 51 dpa. The irradiation data were obtained concerning fuel-cladding chemical interaction, dimensional stability under irradiation and so on. These results showed the superior irradiation performance of the ODS ferritic steel claddings and their application possibility as fast reactor fuels. This paper describes the evaluation of the obtained irradiation data of ODS ferritic steel claddings.

Journal Articles

Effects of neutron irradiation on tensile properties of oxide dispersion strengthened (ODS) steel claddings

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:14 Percentile:76.2(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.

Journal Articles

Oxide fuel fabrication technology development of the FaCT project, 5; Current status on 9Cr-ODS steel cladding development for high burn-up fast reactor fuel

Otsuka, Satoshi; Kaito, Takeji; Yano, Yasuhide; Yamashita, Shinichiro; Ogawa, Ryuichiro; Uwaba, Tomoyuki; Koyama, Shinichi; Tanaka, Kenya

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 6 Pages, 2011/12

This paper describes evaluation results of in-reactor integrity of 9Cr and 12Cr-ODS steel cladding tubes and the plan for reliability improvement in homogeneous tube production. A fuel assembly in the BOR-60 irradiation test including 9Cr and 12Cr-ODS fuel pins has achieved the highest burn-up, i.e. peak burn-up of 11.9at% and peak neutron dose of 51dpa, without any fuel pin rupture and microstructure instability. In another fuel assembly containing 9Cr and 12Cr-ODS steel fuel pins whose peak burn-up was 10.5at%, one 9Cr-ODS steel fuel pin failed near the upper end of the fuel column. A peculiar microstructure change occurred in the vicinity of the ruptured area. The primary cause of this fuel pin rupture and microstructure change was shown to be the presence of metallic Cr inclusions in the 9Cr-ODS steel tube, which had passed an ultrasonic inspection test for defects. In the next stage from 2011 to 2013, the fabrication technology of full pre-alloy 9Cr-ODS steel cladding tube will be developed.

Oral presentation

Charpy impact properties of the irradiated oxide dispersion strengthed (ODS) martensitic steel

Ogawa, Ryuichiro; Yoshitake, Tsunemitsu; Onose, Shoji; Matsumoto, Shinichiro; Sato, Shigeru*

no journal, , 

no abstracts in English

Oral presentation

The Evaluation of strength properties of irradiated PNC316 cladding tube

Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

Transient burst techniques and results of the examination for irradiated PNC316 steel

Nishinoiri, Kenji; Akasaka, Naoaki; Ogawa, Ryuichiro; Inoue, Toshihiko

no journal, , 

In fast reactor, deformation behavior and failure strength of fuel cladding tube (C/T) under loss of coolant flow (LOF) event are important evaluation items of reactor safeties. To evaluate C/T behavior under the primary phase of LOF event, transient bust examination was conducted by neutron irradiated C/T. Specimens of C/T made of PNC316 were irradiated in experimental fast reactor JOYO. In this paper reported the transient burst techniques and the results of the post irradiated examination. In the results, the failure temperature of irradiated C/T has no extreme degradation by comparison of the failure temperature of un-irradiated C/T.

Oral presentation

The Evaluation of tensile strength properties of irradiated PNC1520 cladding tube

Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

Nondestructive evaluation of neutron irradiation damage on austenitic stainless steels by measurement of magnetic flux density

Takaya, Shigeru; Nagae, Yuji; Aoto, Kazumi; Yamagata, Ichiro; Ichikawa, Shoichi; Konno, Shotaro; Ogawa, Ryuichiro; Wakai, Eiichi

no journal, , 

Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01-30 displacement per atom (dpa), 1.0-17 appm and 470-560 $$^{circ}$$C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.

Oral presentation

Evaluation of neutron irradiation damage based on magnetic properties

Takaya, Shigeru; Yamagata, Ichiro; Konno, Shotaro; Ichikawa, Shoichi; Ogawa, Ryuichiro; Nagae, Yuji

no journal, , 

We measured the magnetic flux densities and the magnetization curves on neutron irradiated fast reactor grade type 316 stainless steels by a flux gate sensor and a newly developed vibrating sample magnetometer, respectively. As the result, it was revealed that there is a good relationship between magnetic property and dose, one of representative irradiation damage parameters. This result shows the possibility of nondestructive evaluation of neutron irradiation damage based on magnetic properties.

Oral presentation

Improvement of magnetic flux density measurement technique for irradiation damage evaluation

Konno, Shotaro; Takaya, Shigeru; Nagae, Yuji; Yamagata, Ichiro; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji

no journal, , 

We are developing a method for evaluation of irradiation damage of structural materials in nuclear plants by using change in magnetic flux density due to irradiation damage. In this study, the magnetic flux density measurement technique has been improved by introducing a new magnetizer which enables local magnetizing by contacting the sample surface. We can magnetize samples, especially ferromagnetic samples, more precisely compared to the existing method. Furthermore, the new method provided the path for the application to real plants.

Oral presentation

The Evaluation of strength properties of irradiated PNC1520 cladding tube

Inoue, Toshihiko; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

Improvement of magnetic flux density measurement technique for irradiation damage evaluation

Konno, Shotaro; Takaya, Shigeru; Nagae, Yuji; Yamagata, Ichiro; Ogawa, Ryuichiro; Akasaka, Naoaki; Nishinoiri, Kenji

no journal, , 

We are developing a method for evaluation of irradiation damage on structural materials in nuclear plants by using change in magnetic flux density due to irradiation damage. In this study, the magnetic flux density measurement technique has been improved by introducing a new magnetizer which enables local magnetizing by contacting the sample surface, and the calibration method of the flux gate sensor for the magnetic flux density. We can magnetize samples, especially ferromagnetic samples, more precisely compared to the existing method. Furthermore, the new method can remove many limitations for the application to real plants.

Oral presentation

The Evaluation of heating rate dependency in the transient burst examination of un-irradiated PNC316 and 9Cr-ODS stainless steel cladding tube

Inoue, Toshihiko; Ogawa, Ryuichiro; Inoue, Masaki; Yoshitake, Tsunemitsu; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

Development of ODS ferritic steel claddings for fast reactor fuels, 2; Material irradiation test in JOYO

Yamashita, Shinichiro; Yano, Yasuhide; Ogawa, Ryuichiro; Inoue, Masaki; Yoshitake, Tsunemitsu

no journal, , 

Oxide dispersion strengthened (ODS) steel is a prospective cladding material for the advanced fuel claddings of fast reactors, owing to their excellent radiation resistance and high temperature strength capability. JAEA has been developed two types of 9Cr martensitic and 12Cr ferritic ODS steel claddings, and conducted the irradiation test for the accumulation of irradiation data as well as for the understanding of irradiation behavior of ODS steel claddings. In this study, post irradiation examination data on metallurgical examination, ring-tensile test, hardness measurement, microstructural observation, and chemical analysis were obtained, indicating that there were no significant degradations in mechanical property and also no changes in microstructure due to irradiation and that ODS steel claddings had good irradiation torelance.

Oral presentation

Neutron irradiation effect on mechanical properties of fast breeder reactor grade 316 stainless steel

Takaya, Shigeru; Nagae, Yuji; Ogawa, Ryuichiro

no journal, , 

Neutron irradiation effect on mechanical properties such as tensile and creep was investigated on fast breeder reactor (FBR) grade 316 stainless steel which is one of candidate structural materials for FBR. As a result, it was shown that neutron irradiation effect on mechanical properties of 316FR is similar to that of SUS304.

Oral presentation

Neutron irradiation effects on the tensile properties of FBR structural materials

Takaya, Shigeru; Nagae, Yuji; Ogawa, Ryuichiro

no journal, , 

Neutron irradiation effects on the tensile properties of base metal and weld metal of 316FR is investigated. The 0.2% proof strength of the base metal increased with accumulated fast neutron fluence, while the uniform elongation and the fracture elongation decreased. The tensile strength changed little. These results coincided with the general trends of SUS304 stainless steel. Similar effects are observed for the weld metals.

Oral presentation

Immersion tests of irradiated Zircaloy-2 specimens in artificial seawater

Hayashi, Takehiro; Sasaki, Shinji; Mashiko, Shinichi; Yamagata, Ichiro; Ogawa, Ryuichiro; Inoue, Masaki; Yamashita, Shinichiro; Maeda, Koji

no journal, , 

no abstracts in English

23 (Records 1-20 displayed on this page)