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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.
Isobe, Nobuhiro*; Kawasaki, Nobuchika; Ando, Masanori; Sukekawa, Masayuki*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
Evaluation of local strain at structural discontinuities is an important technology in high temperature design of fast reactors because the failure mode in high temperature fatigue or creep fatigue damage is usually crack initiation and growth from such a locally high strained area. A rationalized strain concentration evaluation method was discussed experimentally in this study. The stress redistribution locus (SRL) method had been proposed to improve the accuracy of local stress and strain evaluation for structural discontinuities. High temperature fatigue tests of circumferentially notched specimens were conducted accompanying with local strain measurement by a capacitance type strain gage. Measured strain was compared with the prediction by the SRL method and the applicability of the method is discussed.
Mayorshin, A. A.*; Skiba, O. V.*; Bychkov, A. V.*; Kisly, V. A.*; Shishalov, O. V.*; Krukov, F. N.*; Novoselov, A. E.*; Markov, D. V.*; Green, P. I.*; Funada, Toshio; et al.
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
The paper presents progress results, including fabrication of vibropac MOX fuel pins and 21 FAs for fast reactor BN-600, irradiation parameters and PIE results. It is shown, that no violations of safe operation limits take place. The activities within the framework of the Demonstration experiment is based on the international cooperation and have been performed with the support and participation Russian and Japanese organizations RIAR, IPPE, OKBM, BNPP, MEXT, JAEA, PESCO. The goal of the experiment is to validate possibility of using vibropac MOX FA for weapon plutonium disposition.
Aizawa, Kosuke; Oshima, Jun*; Kamide, Hideki; Kasahara, Naoto
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product or delayed neutron. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. To overcome above diffculties, we have developed the sampling method for indentifying the failed fuel subassemblies located under the slit by numerical simulations and water experiments.
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.
Kurome, Kazuya*; Kawamura, Masaya*; Enuma, Yasuhiro*; Tsujita, Yoshihiro*; Sato, Mitsuru*; Futagami, Satoshi; Hayafune, Hiroki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 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.
C core outlet temperatureNaganuma, Masayuki; Ogata, Takanari*; Mizuno, Tomoyasu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
In the FaCT project, a design study of the metallic fuel SFR has been executed as secondary candidate. The primary interest is to achieve a core outlet temperature of 550 C. However, the metallic fuel has a drawback that the maximum temperature of the cladding inner surface is limited to 650 C to avoid liquid phase formation. To overcome this problem, JAEA has developed and studied the advanced core concept with single Pu-enrichment and 2 radial regions of heavy metal density. In this paper, the core and fuel design study for the middle-scale SFR applying this core concept are discussed. In addition, for the practical application of the metallic fuel to the SFR with high outlet temperature, it is necessary to expand irradiation experience under the high cladding temperature condition. Therefore, JAEA and CRIEPI planned an irradiation test of the metallic fuel in Joyo as a collaborative program. In this paper, the outline and current status of the irradiation test are reported.
Kawashima, Katsuyuki; Ogawa, Takashi; Oki, Shigeo; Okubo, Tsutomu; Mizuno, Tomoyasu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00
Sodium-cooled fast reactor core design considerations are made to improve the proliferation resistance by focusing on the plutonium generated in the UO
blanket in the frame of the Fast Reactor Cycle Technology Development (FaCT) project. The appropriate design and treatments of the UO blanket help to reduce the intrinsic proliferation potentials. Based on the 1500 MWe FaCT reference core, the three different cores (radial blanket-free core, the core with the low-enriched MOX fuel, and the core with MA-doped UO fuel) are configured to meet the provisional proliferation resistance criteria as well as the core performance targets.
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.
Aoto, Kazumi; Uto, Nariaki; Sakamoto, Yoshihiko; Ito, Takaya*; Toda, Mikio*; Kotake, Shoji*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00
In the FaCT project, SFR with 1,500 MWe is a target for the commercialization. R&D on innovative technologies to achieve the economic competitiveness and enhance the reliability and safety is carried out. A compact RV without wall-cooling layer is pursued in consideration of seismic reliability. For a two-loop cooling system with shortened high chromium steel piping, studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being performed. A double-walled straight tube SG is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing including the thermal-hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR and experimental demonstration in JOYO. An advanced fuel handling system is also pursued. Discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010.
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.
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.
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
