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Nakai, Ryodai
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009), p.207 - 220, 2012/00
The concept of defense in depth (DiD) shall be applied to the safety design of advanced SFRs. Through the prevention, detection and control of accident, core disruptive accident (CDA) shall be excluded from Design Basis Events (DBEs). Considering that the SFR reactor core is not the most reactive configuration unlike in the LWRs, design measures to prevent and mitigate the consequences of CDA are being considered as provisions for beyond design basis events (BDBEs). To effectively meet the future nuclear energy system safety goals, advanced SFR designs should exploit passive safety features to increase safety margins and to enhance reliability. In particular the safety approach to eliminate the severe re-criticality will be highly desirable, because with this approach, severe accidents in SFRs can be simply regarded as similar to LWRs.
Okamura, Shigeki*; Eto, Masao*; Kamishima, Yoshio*; Negishi, Kazuo; Sakamoto, Yoshihiko; Kitamura, Seiji; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
This paper describes the seismic design of JSFR, which includes the seismic condition, the seismic isolation system and the seismic evaluation of primary component. JSFR employs a seismic isolation system to mitigate the earthquake force. The design seismic loading is made more severe than ever since Niigata-ken Chuetsu-oki Earthquake in 2007. The earthquake force loaded on the primary components has to be mitigated more than that of the previous seismic isolation system. We examined the advanced seismic isolation system by optimizing the performance of the previous seismic isolation system considering the natural frequency of the primary components. The advanced seismic isolation system for SFR was adopted laminated rubber bearings which are thicker than that of the previous, as well as oil dampers. The seismic evaluation of nuclear reactor components under applying the advanced seismic isolation system was performed; the performance of the system was confirmed.
D'hondt, P. J.*; Yamagishi, Isao; Weaver, D. R.*; Nordborg, C.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 15 Pages, 2012/00
The OECD/NEA has recently published the report on Research and Test Facilities Required in Nuclear Science and Technology, as well as a database listing facilities (http://www.nea.fr/rtfdb/). The paper focuses on recommendations which relate to the status and anticipated developments in Fast Reactor technology in the medium term and makes comments on the needs for zero and low-power reactors and sub-critical assemblies to extend the knowledge of the skills base as well as research reactors and critical assemblies related to particular reactor designs. The paper emphasises the recommendation; further federation of the financial, scientific efforts of the OECD-countries to optimise available resources, the key role played by international institutions, the maintenance of the current skills base of researchers. The paper emphasises the issue of knowledge retention, for example through databases of older experiments such as IRPhE.
Kato, Atsushi; Kotake, Shoji; Yoshiuji, Takahiro*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
Within the FaCT project, commodities shall be reduced by introducing innovative technologies. In order to evaluate the economy for the Japan Sodium-cooled Fast Reactor (JSFR), the account code named SCALLE (Sum of Cost Account Leading to future Logistics Economy) has been developed, in which the basic methodology is bottom up of component costs based on amounts of material and corresponding unit costs.
Kato, Atsushi; Chikazawa, Yoshitaka; Uto, Nariaki; Hirata, Shingo; Obata, Hiroyuki*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
A basic feasibility of the helium gas cask has been evaluated by thermal analyses. There have been conducted two analyses: whole cask and detail inside subassembly analyses. The detail inside subassembly analysis has shown that the temperature distribution is mainly governed by thermal conductivity and natural convection of coolant helium hardly contributes heat removal. In the case of a cask with five subassemblies with 2.2 kW decay heat per each, the maximum cladding temperature is evaluated to be 361 
C satisfying cladding temperature limit of 395 C. Those results have shown the basic feasibility of the helium gas fresh fuel shipping cask.
Nakamura, Kinya*; Kato, Tetsuya*; Ogata, Takanari*; Nakajima, Kunihisa; Iwai, Takashi; Arai, Yasuo
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
The first irradiation campaign of U-Pu-Zr metal fuel in Japan is planned in the experimental fast reactor JOYO. In the fabrication of U-Pu-Zr fuel, two methods were adopted for preparing U-Pu alloy from the oxide; one is the electrochemical reduction and the other is the electrorefining followed by reductive extraction. Injection casting for U-Pu-Zr slug was carried out after adding U and Zr metals to meet the target specifications of the irradiated fuel. Several conditions of Na-bonding process were determined from the results of tests using simulated metal fuel pins. Based on these results, six U-Pu-Zr fuel pins for the irradiation tests are now being fabricated.
Kuno, Yusuke; Senzaki, Masao; Seya, Michio; Inoue, Naoko
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
A large amount of plutonium as well as high 
Pu should be handled in the future fast reactor nuclear fuel cycle (FR-NFC), where very robust measures for nuclear proliferation-resistance (PR) may have to be taken to prevent nuclear proliferation. To find a good balance of extrinsic barrier and intrinsic one will come to be essential for NFC designers to optimize civilian nuclear technology with nuclear non-proliferation. International Safeguards including Comprehensive Safeguards Agreement and Additional Protocol is the most effective institutional barrier among other institutional measures in non-proliferation regime. The advanced Safeguards with high detectability can play a dominant role for PR in the states complying with full institutional controls. In this context, a new concept of differentiation in the intrinsic measures depending upon the level of Safeguards could be applied from the viewpoint of plant design rationalization.
Ito, Chikara; Araki, Yoshio; Naito, Hiroyuki; Iwata, Yoshihiro; Aoyama, Takafumi
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
no abstracts in English
Chikazawa, Yoshitaka; Kotake, Shoji; Sawada, Shusaku*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
experiments on the flow and freezing of metal fuel and cladding melts, 1; Test conditions and overview of the resultsFukano, Yoshitaka; Kawada, Kenichi; Sato, Ikken; Wright, A. E.*; Kilsdonk, D. J.*; Aeschlimann, R. W.*; Bauer, T. H.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
experiments on the flow and freezing of metal fuel and cladding melts, 2; Results, analysis, and applicationsWright, A. E.*; Bauer, T. H.*; Kilsdonk, D. J.*; Aeschlimann, R. W.*; Fukano, Yoshitaka; Kawada, Kenichi; Sato, Ikken
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00
Tsukimori, Kazuyuki; Ueda, Masashi; Miyahara, Shinya; Yamashita, Takuya
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
Hirata, Shingo; Chikazawa, Yoshitaka; Kato, Atsushi; Uto, Nariaki; Obata, Hiroyuki*; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In Fast Reactor Cycle Technology Development (FaCT) project, Japan Sodium-cooled Fast Reactor (JSFR) is going to adopt an advanced fuel handling system. As for ex-vessel spent fuel handling, a pot which contains two fuel subassemblies simultaneously and is applicable size to compact reactor vessel, has been developing so as to shorten a refueling period leading to an improvement of plant availability. The pot is required to provide sufficient cooling capability even in case of transportation malfunction during transportation of spent fuel subassemblies with high decay heat. In the present study, experimental and analytical studies to evaluate the cooling capacity of the pot are summarized.
Sawada, Makoto; Sasaki, Kazuichi; Nishida, Masaaki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
Japan is aiming at starting commercial operation of a demonstration FBR around 2025 in the FaCT (Fast Reactor Cycle Technology Development) project. To prepare for such as the new FBR age, INITC has established a total of 27 JAEA staff training courses based on the teachings obtained from the Monju leak accident, regarding to FBR operation technology, sodium handling technology, maintenance technology and FBR plant system engineering technology, and also has been conducting energy environmental education for from under high school students to graduate students of the whole country including local universities. In addition, INITC aims become a central of excellent (COE) of the international technology training in Asia through the international educational training programs sponsored by MEXT. The variety of the activities of educational training mentioned above will contribute to the development of the human resource in Japan and abroad, towards the next generation age.
Nakatsuka, Toru; Misawa, Takeharu; Yoshida, Hiroyuki; Takase, Kazuyuki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 5 Pages, 2012/00
In the present paper, thermal-hydraulic behavior in a simplified fuel assembly of the supercritical water cooled fast reactor (Super Fast Reactor) was analyzed with the three-dimensional two-fluid model analysis code ACE-3D. The analytical geometry simulates a 19-rod assembly, which is one of the most simplified geometry of the SCWR fuel assembly and includes three kinds of different subchannel types; (1) adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). It was confirmed that the MCST satisfies a thermal design criteria to ensure fuel and cladding integrity.
Kubo, Shigenobu*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kotake, Shoji
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
In order to embody a safety design, a higher level safety principle was broken down into a set of design requirements for each safety related system, structure and component (SSC). This paper will present an output of the safety requirements for safety related SSCs of JSFR.
Hamada, Noriaki*; Shiina, Koji*; Fujimata, Kazuhiro*; Hayakawa, Satoshi*; Watanabe, Osamu*; Yamano, Hidemasa
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In a sodium-cooled fast reactor, a vortex cavitation evaluation methodology was developed to predict a possible cavitation generated by vortex at the center of accelerating flow. This methodology was applied to a scaled model experiment, leading to the prospect that the cavitation can be predicted.
Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Murakami, Takahiro*; Iwamoto, Yukiharu*; Yuki, Kazuhisa*; Sago, Hiromi*; Hayakawa, Satoshi*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in JSFR, in particular emphasizing on recent R&D activities that investigate unsteady elbow pipe flow. The experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was distorted at the downstream from the elbow, the experiment clarified that the effect of swirl flow on pressure fluctuation onto the pipe wall was not significant. The simulation revealed that Reynolds number scarcely affects flow patterns and flow velocity distributions.
Eto, Masao*; Kamishima, Yoshio*; Okamura, Shigeki*; Watanabe, Osamu*; Oyama, Kazuhiro*; Negishi, Kazuo; Kotake, Shoji*; Sakamoto, Yoshihiko; Kamide, Hideki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00
In the JSFR design, the diameter of the Reactor Vessel (RV) shall be minimized and the reactor internal structures shall be simplified for reduction in construction cost. The reduction in the RV diameter is achieved by adopting an advanced refueling system and the hot RV with high temperature wall. The flow velocity in the reactor upper plenum increases because the diameter of the RV is decreased. As the result, the coolant flow field in reactor upper plenum is severe. The optimization of the coolant flow field in the reactor upper plenum was carried out for prevention the cover gas entrainment and the vortex cavitations at the hot leg intake. In addition, structural integrities for seismic loadings and thermal loadings were evaluated because the design seismic loading was highly increased and the vessel wall is directly exposed to the thermal transients of the upper plenum. This paper describes the characteristics and the results of the design study of the reactor system.
Konomura, Mamoru; Ichimiya, Masakazu; Mukai, Kazuo
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00
Japanese prototype fast breeder reactor, Monju, will soon restart. Monju has three sodium loops with steam generators and a turbin-generator, 280 MWe. Monju is expected to demonstrate the sodium technology and the sodium-water heat exchange technology in Japan. It will be operated at full power operation during around 10 years after restart in order to accumulate operation & maintenance experience and to evaluate its design technology. After the system start-up test (SST), Monju will be operated under full power. In this stage, the main object of Monju operation will be to achieve its initial targets which were fixed when its construction was decided around thirty years ago. The targets are to demonstrate a safe and reliable operation, that is, accumulation of the operation & maintenance experience and evaluation of the design technology, and to establish sodium handling technology. For example, inspection and diagnosis technologies are important for maintenance of a sodium cooled reactor. An out-of-pile sodium test facility will be constructed near Monju in order to make many tests of inspection devices and research many chemical tests. At the same time, the activities for the performance improvement, for example, a new licence, will be prepared in order to utilize Monju as a R&D facility. After the accumulation of operating experience, Monju will be enhanced the performance as a R&D facility in order to demonstrate innovative technologies, for example, irradiation of advanced fuel, longer operation cycle, higher burnup. For this purpose, Monju will be needed to get a new licence and core modification. And Monju on-site non-destructive Post Irradiated Evaluation facility will be expected at this stage. There were many R&D works in Japan with sodium out-of-pile facilities. All the experience were reflected in the design of Monju. Monju will demonstrate a handling of sodium technologies under power plant operation.
