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Journal Articles

Evaluation on coolability of the reactor core in Monju by natural circulation under earthquake and subsequent tsunami event

Yamada, Fumiaki; Fukano, Yoshitaka; Nishi, Hiroshi; Konomura, Mamoru

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

The core cooling capability by natural circulations at a station black-out event, induced by an earthquake and a subsequent tsunami attack, has been evaluated in detail, referring to the accident of the Fukushima Dai-ichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code: Super-COPD has been used for the evaluation, which has been verified by the analyses of the preliminary test results on the natural circulation in Monju. As a result it was concluded that the natural circulations of the coolant sodium will enable the decay heat removal of the core as far as the sodium coolant flow circuits are intact and secured.

Journal Articles

Restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor "Joyo", 2

Takamatsu, Misao; Ashida, Takashi; Kobayashi, Tetsuhiko; Kawahara, Hirotaka; Ito, Hideaki; Nagai, Akinori

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

In the experimental fast reactor Joyo, it was confirmed that the top of the irradiation test Sub-Assembly (S/A) of "MARICO-2" (material testing rig with temperature control) had bent onto the in-vessel storage rack (IVS) as an obstacle and had damaged the Upper Core Structure (UCS). This incident necessitates the replacement of the UCS and the retrieval of MARICO-2 S/A for Joyo re-start. Along with four stages involving jack-up and retrieval of the existing damaged UCS (ed-UCS), retrieval of the MARICO-2 S/A, and installation of the new UCS (n-UCS) in the restoration work plan, current conditions at Joyo are being carefully investigated, and the results are applied to the designs of special handling equipment, which is being manufactured and scheduled for operation in 2014.

Journal Articles

Evaluation of feedback reactivity in Monju start-up test

Kitano, Akihiro; Miyagawa, Takayuki*; Okawachi, Yasushi; Hazama, Taira

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03

The feedback reactivity was measured in Monju start-up test conducted in 2010. The two reactivity components related either to power or to the core inlet coolant temperature were evaluated by fitting to a reactivity balance equation as a function of neutron count rate and coolant temperature. The measured feedback reactivity and the two components were compared with calculation taking account of the temperature distribution in the core. The calculated and the measured values of the feedback reactivity showed a reasonable agreement.

Journal Articles

Oxidation and reduction behaviors of plutonium and uranium mixed oxide powders

Hiroka, Shun; Kato, Masato; Tamura, Tetsuya*; Nelson, A. T.*; McClellan, K. J.*; Suzuki, Kiichi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 8 Pages, 2013/03

As research and development activities for MOX fuel pellet production, oxidation and reduction behaviors of MOX powders were investigated by thermogravimetry and X-ray diffraction measurements. It was observed that the oxidation limit decreased with oxidizing temperature and Pu content. The MOX powders showed a two-step oxidation and kinetic stability under non-stoichiometry. The oxidation rates were evaluated from the isothermal oxidation tests. It was found that the reduction temperature of M$$_{4}$$O$$_{9}$$ + M$$_{3}$$O$$_{8}$$ was higher than that of M$$_{4}$$O$$_{9}$$. This indicated that the reduction of M$$_{4}$$O$$_{9}$$ was prevented by the existence of M$$_{3}$$O$$_{8}$$. Activation energy of the reduction was derived from the non-isothermal reduction tests. The data are expected to contribute to establishing a control technique for O/M ratio during MOX powder storage and pellet production.

Journal Articles

Sinterability of ZrN and (Zr$$_{0.6}$$Dy$$_{0.4}$$)N pellets; Surrogate fuel fabrication for ELECTRA

Pukari, M.*; Takano, Masahide

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 8 Pages, 2013/03

Pellets of inert matrix material ZrN and surrogate nitride fuel material (Zr$$_{0.6}$$Dy$$_{0.4}$$)N are fabricated for the purpose of investigating the origin and the effect of C and O impurity concentrations. Oxygen concentrations of up to 1.5 wt% were deliberately introduced to the materials with two separate methods. The achievable green and sintered pellet densities of these materials as a function of O content, dimensional properties of the powder and sintering temperature are shown. The effect of O dissolved into the matrix vs oxide-rich phase segregation is discussed. Oxygen pickup during the fabrication of the product as well as its exposure to air is demonstrated. The quality of the materials is monitored by the systematic analysis of O, N and C contents throughout the fabrication and sintering processes, supported by XRD and SEM analyses.

Journal Articles

Information sharing framework among experts for facilitating development of fast reactors and fuel cycles

Kawakubo, Yoko; Hoffheins, B.; Inoue, Naoko; Mongiello, R.*; Baldwin, G.*; Lee, N. Y.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 8 Pages, 2013/03

Transparency in the peaceful use of nuclear energy is important as a measure to complement and reinforce International Atomic Energy Agency (IAEA) safeguards and promote international/ regional confidence building. Moreover, information sharing, a key component of confidence building, is essential for promoting the development of fast reactors and associated fuel cycles by enhancing transparency and encouraging understanding among non-proliferation experts. Currently, Japan Atomic Energy Agency (JAEA) is carrying out a joint project to design and establish an Information-Sharing Framework (ISF) for supporting and promoting nuclear transparency in the Asia Pacific region, in cooperation with Sandia National Laboratories (SNL), the Korean Institute for Nonproliferation and Control (KINAC), and Korea Atomic Energy Research Institute (KAERI). At present, requirements for planning and implementing ISF are under discussion to address inherent challenges that are recognized among project partners. This paper describes the current status of the development of requirements for ISF. The effort of the development is still underway, however; the requirements will be determined and demonstrated in the near future by project partners.

Journal Articles

Evaluation on calculation accuracy of the sodium void reactivity for low void effect fast reactor cores with experimental analyses

Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03

Calculation accuracy of the sodium void reactivity for safety-enhanced fast reactor core concepts was evaluated with analyses of critical experiments. In these concepts, heterogeneous core configuration and sodium plenum replacement are adopted to reduce the sodium void reactivity to around zero. In the past, a variety of critical experiments for heterogeneous cores had been carried out in the ZPPR facility, some of which are compiled in the IRPhEP handbook. Further, several experiments for core with sodium plenum had been performed in the BFS-2 facility. Calculation analyses of above mentioned critical experiments have been performed by using the Japanese current reactor physics analytical system. These analyses clarified following items: (1) Accuracy for the axially-heterogeneous core was comparative or less to that of the homogeneous core. However, accuracy for the radially-heterogeneous core was not satisfactory. (2) Accuracy for the core with sodium plenum was not satisfactory in the sodium plenum voiding case.

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

Japanese FR deployment scenario study after the Fukushima accident

Ono, Kiyoshi; Shiotani, Hiroki; Ohtaki, Akira; Mukaida, Kyoko; Abe, Tomoyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

In parallel to the mid-term analyses by the AEC after the Fukushima-accident, JAEA implemented the long-term scenario analyses for the nuclear fuel cycle options including FR cycle deployment. As a result, it was revealed that FR cycle deployment brings great benefits to reduction of uranium demand, spent fuel storage, radioactive waste generation, and Pu stockpiles in addition to potential hazard of HLW in "20 GWe constant after 2030" case. Meanwhile, it was also revealed the benefits of reduction of radioactive waste generation and Pu stockpiles in "Gradual decrease from 20 GWe after 2030" case.

Journal Articles

Enhancement of JSFR safety design and criteria for Gen.IV reactor

Aoto, Kazumi; Chikazawa, Yoshitaka; Okubo, Tsutomu; Okada, Keizo*; Ito, Takaya*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

Overview of Japan Sodium-cooled Fast Reactor (JSFR) development status and reflection of lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plant (1F) accident have been summarized. JSFR was recognized as a promising next generation nuclear reactor. Even though the JSFR safety design already took into account measures against severe accident situations and passive safety features such as passive shutdown system and natural convection decay heat removal systems in the 2010 design version, it is become aware of importance of design measures against severe accidents and extreme external events by the 1F accident. As recent activities, external hazard evaluations and design improvements reflecting lessons learned from 1F accident have been conducted. This paper also discusses importance of development of global safety design criteria and international Research and Development cooperation on safety design measures.

Journal Articles

Evaluation of severe external events on JSFR

Hayafune, Hiroki; Kato, Atsushi; Chikazawa, Yoshitaka; Okubo, Tsutomu; Sagawa, Hiroshi*; Shimakawa, Yoshio*

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

Evaluation of earthquake and tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection.

Journal Articles

Safety design approach for JSFR toward the realization of GEN IV sodium cooled fast reactor

Kubo, Shigenobu; Yamano, Hidemasa; Chikazawa, Yoshitaka; Shimakawa, Yoshio*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03

This paper describes the safety design approach for JSFR. To achieve safety goals for Generation IV reactor, design measures should be enhanced against design extension conditions including those for external events considering the lessons learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident. The current safety design approach for JSFR intends to meet the safety design criteria for Generation-IV SFR developed in the framework of the Generation-IV International Forum. Design extension conditions and related design measures are identified and selected with due consideration of the safety features of SFR. Design approach and measures for severe external events such as earthquake and tsunami, external missiles, failure to shutdown type events and failure to heat removal type events are shown. Several situations to be practically eliminated are proposed with possible design measures.

Journal Articles

Recent progress and status of Monju

Kondo, Satoru; Deshimaru, Takehide; Konomura, Mamoru

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 6 Pages, 2013/03

Monju, a 280-MWe prototype sodium-cooled fast reactor of, restarted its test operation in 2010. The zero-power tests were successfully conducted. It's major achievement was accurate prediction of reactor physics parameters with a core including americium-rich fuel. The reactor, however, has been put into a stand-by mode again since the 3.11 Fukushima-Daiichi accident. The roles shall not change: demonstrating stable power generation and actinide burning; providing technology and knowledge base for future SFRs; and using the plant as an international research facility.

Journal Articles

Cooperation on impingement wastage experiment of Mod. 9Cr-1Mo steel using SWAT-1R sodium-water reaction test facility

Beauchamp, F.*; Nishimura, Masahiro; Umeda, Ryota; Allou, A.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 10 Pages, 2013/03

T91 is one of the material candidates of SGU tubes for future sodium-cooled fast reactors (SFRs). Wastage characterization of T91 is needed to evaluate the consequences for safety and the availability of the SGU. Six T91 target tubes were incorporated in the SWR test facility (SWAT-1R) of JAEA and subjected to reaction jets. All tubes were successfully penetrated by the reaction jets, and the wastage rates were determined. This paper describes the SWAT-1R facility, the test procedure and operating conditions, and brings out the main results and experience gained through the wastage experiments.

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