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JAEA Reports

BDI behavior evaluation of an upgraded Monju core and a demonstration core, 1; Plans for the out of pile bundle compressive tests for large diameter pins

Ichikawa, Shoichi; Haga, Hiroyuki; Katsuyama, Kozo; Uwaba, Tomoyuki; Maeda, Koji; Nishinoiri, Kenji

JAEA-Testing 2012-001, 36 Pages, 2012/07

JAEA-Testing-2012-001.pdf:11.48MB

The life of FBR (Fast Breeder Reactor) fuel assembly is restricted by BDI (Bundle-Duct Interaction). Therefore, it is very important to carry out the out pile bundle compressive tests which can imitate BDI, in order to evaluate BDI behavior. The target of the conventional BDI behavior was thin pins ($$phi$$ 6.5 mm) for fuel pellets which were used with the assembly of Monju (the Monju prototype fast breeder reactor) etc. Furthermore by an upgraded Monju core and a demonstration core, adoption of thick pins for the holler annular pellets is planned. Therefore, it was necessary to carry out BDI evaluation of a thick pin. Then, the plans for out of pile bundle compressive test for large diameter pins were are reported this time.

JAEA Reports

Establishment of reassembly technique of capsule type irradiation rig

Ichikawa, Shoichi; Abe, Kazuyuki; Haga, Hiroyuki; Kajima, Hisashi*; Sakurai, Satoshi*; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

JAEA-Technology 2011-032, 46 Pages, 2012/01

JAEA-Technology-2011-032.pdf:8.46MB

The assembly technique to the capsular irradiation rig newly developed was established. In the irradiation examination, the assembling disassembling and reassembling to PFB110 "B11(1), B11(2)" and PFB140 "B14" that built in Am-MOX fuel pin was achieved. The reassembly technique by recycling the irradiation material was established in the assembly of B11(2). This time, the assembly and disassembly of B11 (1) were reported. Moreover, the assembly of B14 which improved the assembly technology of B11 (1) was reported.

Journal Articles

Interrelationship between true stress-true strain behavior and deformation microstructure in the plastic deformation of neutron-irradiated or work-hardened austenitic stainless steel

Kondo, Keietsu; Miwa, Yukio; Tsukada, Takashi; Yamashita, Shinichiro; Nishinoiri, Kenji

Journal of ASTM International (Internet), 7(1), p.220 - 237, 2010/01

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

Irradiation effect on microstructure of modified SUS316 stainless steel cladding irradiated at elevated temperature to high dose

Yamashita, Shinichiro; Akasaka, Naoaki; Nishinoiri, Kenji; Takahashi, Heishichiro

no journal, , 

A modified SUS316 stainless steel (PNC316) for high burn-up fast reactor core application has been successfully developed. The material has a typical composition of 16Cr-14Ni-2.5Mo-0.25P-0.004B-0.1Ti-0.1Nb and is used in the 20% cold-worked condition. To demonstrate irradiation performance of PNC316 cladding irradiation experiments has been successively carried out at the temperature ranging from 775 K to 905 K up to a maximum dose of 125 dpa. In this study, the microstructures of the irradiated specimens were carefully examined. Low-magnification image including some of needle-like precipitates indicates many helium bubbles attaching at the precipitate interface. On the other hand, a representative high-resolution image of the needle-like precipitate with the sizes of a few tens of nanometers long and a few nanometers wide shows a distinct interface structure between FCC matrix and the precipitate. As the results of diffraction pattern and image analyses this characteristic precipitate was identified as Fe$$_{2}$$P phase. Based on the numerous results of TEM examination, it is found that a significant improvement in the swelling resistance of PNC316 was mainly derived from the formation of a stable phosphide that traps helium bubbles and retards the conversion of bubbles to voids.

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

Effect of fine precipitate on swelling behavior of heavily-irradiated PNC316 stainless steel

Yamashita, Shinichiro; Sekine, Manabu*; Tachi, Yoshiaki; Akasaka, Naoaki; Nishinoiri, Kenji; Takahashi, Heishichiro

no journal, , 

no abstracts in English

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

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

Effect of nature of grain boundary on void distribution during irradiation in austenitic stainless steels

Sekio, Yoshihiro; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Takahashi, Heishichiro*

no journal, , 

In this study, microstructural observation of Fe-15Cr-15Ni ternary model alloy and PNC316 irradiated in JOYO was performed to clarify the defect accumulation process in the vicinity of grain boundary and also the relationship between grain boundary coherency and the void structure developing process was evaluated. As the result of microstructural observation, the followings are obtained; void denuded zone (VDZ) was developed in the vicinity of the low symmetry grain boundary of Fe-15Cr-15Ni whereas no VDZ developed in the vicinity of high symmetry grain boundary. In the case of PNC316, narrow VDZ was developed only at the low symmetry grain boundary.

Oral presentation

Irradiation behavior of oxide dispersion strengthened ferritic steel cladding irradiated in JOYO

Yamashita, Shinichiro; Yano, Yasuhide; Otsuka, Satoshi; Kaito, Takeji; Akasaka, Naoaki; Inoue, Masaki; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Koyama, Shinichi; Tanaka, Kenya

no journal, , 

In this work, neutron irradiation behaviour of ODS ferritic steel cladding tubes developed for fast reactor (FR) was investigated to understand the effect of neutron irradiation on their microstructures. Chemical compositions of the ODS cladding tubes examined were Fe-0.13C-8.84Cr-1.97W-0.20Ti-0.34Y$$_{2}$$O$$_{3}$$(9Cr-ODS) and Fe-0.04C-11.34Cr-1.89W-0.25Ti-0.23Y$$_{2}$$O$$_{3}$$ (12Cr-ODS). These ODS cladding tubes were irradiated, without fuel condition, at 731-1089 K to fast fluences ranging from 3.2 to 6.6$$times$$10$$^{26}$$ n/m$$^{2}$$ (E $$>$$ 0.1 MeV) in the experimental fast reactor JOYO. Microstructural stability of these cladding tubes was evaluated using transmission electron microscope (TEM). Density of the tube specimens before and after irradiation was measured by a conventional immersion method with water, indicating that no significant swelling occurred for all the irradiated specimens. TEM observations show that the radiation-induced defect cluster formation during neutron irradiation was suppressed. It was highly possible due to high density defect sink site such as initially-existed dislocation introduced during tube fabrication process, interface between precipitates including oxide and each matrix. In addition, it revealed that oxide particles, which are closely related with high temperature strength under the practical reactor operation, were stable up to the maximum doses of this irradiation test from the analyses of TEM micrographs.

Oral presentation

Study on the plastic deformation behavior of neutron irradiated austenitic stainless steels

Kondo, Keietsu; Nakano, Junichi; Tsukada, Takashi; Yamashita, Shinichiro; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

Effects of nature of grain boundary on void distribution during irradiation in austenitic stainless steels

Sekio, Yoshihiro; Yamashita, Shinichiro; Yoshitake, Tsunemitsu; Nishinoiri, Kenji; Takahashi, Heishichiro

no journal, , 

The process of the irradiation-induced defects accumulation during irradiation has not been clarified yet, namely what microstructural factors affect defect accumulation process. For this reason, we investigated the void distribution near grain boundaries by focusing on grain boundaries as a microstructural factor in Fe-15Cr-15Ni model alloy and PNC316 stainless steel used for the nuclear fast reactor. It was clarified that the void denuded zone were formed near only a random grain boundary under neutron irradiation in Fe-15Cr-15N alloy and PNC316 stainless steel as being observed in the electron-irradiated austenitic stainless steels. We discussed the difference of the void denuded zone distribution in both steels under low and high fluencie. These results suggest that the grain boundary characteristics and its effect on void denuded zone formation will play a role for the void swelling under neutron irradiation in austenitic stainless steels.

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

Irradiation behavior of annular fuel pellets, 1; PIE results of X-ray CT test

Ishimi, Akihiro; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

no journal, , 

no abstracts in English

Oral presentation

BDI behavior evaluation of "Monju" advanced reactor and demonstration reactor, 1; Out of pile bundle compressive test plan for a thick pin

Ichikawa, Shoichi; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji; Uwaba, Tomoyuki; Koyama, Shinichi

no journal, , 

Out of pile bundle compressive test plan for a thick pin is carried out in order to evaluate BDI behavior of "Monju" advanced reactor and demonstration reactor.

Oral presentation

Evaluation of fission products migration behavior by fuel heating tests, 1; Objective and outline

Miwa, Shuhei; Ishimi, Akihiro; Tanaka, Kosuke; Sato, Isamu; Hirosawa, Takashi; Obayashi, Hiroshi; Koyama, Shinichi; Katsuyama, Kozo; Nishinoiri, Kenji; Osaka, Masahiko; et al.

no journal, , 

The experimental studies for the evaluations of fission products migration behaviors focused on the evaluations of chemical form were planned to improve the evaluation accuracy of the source term in a severe accident of nuclear power plant.

19 (Records 1-19 displayed on this page)
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