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Journal Articles

Thermal-hydraulic calculation for simplified fuel assembly of super fast reactor using two-fluid model analysis code ACE-3D

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.

Journal Articles

Safety design requirements for safety systems and components of JSFR

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.

Journal Articles

Development of computational method for predicting vortex cavitation in the reactor vessel 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.

Journal Articles

Unsteady elbow pipe flow to develop a flow-induced vibration evaluation methodology for JSFR

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.

Journal Articles

Comparison of pool/loop configurations in the JAEA feasibility study 1999-2006

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

Journal Articles

CAF$'E$ experiments on the flow and freezing of metal fuel and cladding melts, 1; Test conditions and overview of the results

Fukano, 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

Journal Articles

CAF$'E$ experiments on the flow and freezing of metal fuel and cladding melts, 2; Results, analysis, and applications

Wright, 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

Journal Articles

Advanced light water reactor with hard neutron spectrum for realizing flexible plutonium utilization (FLWR)

Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 8 Pages, 2012/00

An advanced LWR with hard neutron spectrum named FLWR is a BWR-type reactor with a core consisting of hexagonal-shaped fuel assemblies with a triangular tight-lattice fuel rod configuration. It has been proposed in order to ensure sustainable energy supply in the future based on the well-experienced LWR technologies. The reactor concept of the FLWR is designed to utilize the most of the existing Advanced Boiling Water Reactor (ABWR) plant system. Therefore, only the core concept is new. The FLWR aims at effective and flexible utilization of uranium and plutonium resources by adopting a two-stage concept of core designs. The core in the first stage of FLWR is for intensive utilization and conservation of plutonium with no degradation of the isotopic quality of plutonium based on the experience of the current LWR-MOX utilizations. The one in the second stage realizes sustainable multiple plutonium recycling with a high conversion ratio over 1.0. When the technologies and infrastructures for multiple recycling with MOX spent fuel reprocessing are established, the core of the first stage proceeds to the second stage by only changing the fuel assembly design in the same reactor system. The present paper summarizes the recent core design studies of FLWR.

Journal Articles

Human development in Japan and abroad using prototype FBR Monju towards the next-generation age

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.

Journal Articles

Experimental investigation of strain concentration evaluation based on the stress redistribution locus method

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.

Journal Articles

Progress of demonstration experiment on irradiation of vibro-packed MOX fuel assemblies in the BN-600 reactor

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.

Journal Articles

Experimental and analytical study of failed fuel detection and location system in JSFR

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.

Journal Articles

Development of FR construction cost estimation method in FaCT project

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.

Journal Articles

Thermal analysis on shipping cask for JSFR fresh fuel

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 $$^{circ}$$C satisfying cladding temperature limit of 395 $$^{circ}$$C. Those results have shown the basic feasibility of the helium gas fresh fuel shipping cask.

Journal Articles

R&D on maintenance technologies for FBR plants in JAEA; The Status quo and the future plan

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

Journal Articles

Development of transfer pot for JSFR ex-vessel fuel handling

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.

Journal Articles

U-Pu-Zr metallic fuel core and fuel concept for SFR with a 550$$^{circ}$$C core outlet temperature

Naganuma, 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 $$^{circ}$$C. However, the metallic fuel has a drawback that the maximum temperature of the cladding inner surface is limited to 650 $$^{circ}$$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.

Journal Articles

Fast reactor core design considerations from proliferation resistance aspects

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$$_{2}$$ blanket in the frame of the Fast Reactor Cycle Technology Development (FaCT) project. The appropriate design and treatments of the UO$$_{2}$$ 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$$_{2}$$ fuel) are configured to meet the provisional proliferation resistance criteria as well as the core performance targets.

Journal Articles

U-Pu-Zr metal fuel fabrication for irradiation test at JOYO

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.

Journal Articles

In-service inspection and repair program for commercialized sodium-cooled fast reactor

Nishiyama, Noboru; Kotake, Shoji*; Uzawa, Masayuki*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00

Japan Sodium-cooled Fast Reactor (JSFR) adopts a compact design to reduce construction cost as a commercialized reactor. A commercialized fast reactor is also expected to be excellent in operation and maintenance for high availability. From the study for JSFR, in-service inspection and repair program necessary for JSFR has been developed. Major component design has been improved considering ISI accommodation, and the required ISI devices have been specified according to the proposed ISI program. To increase maintainability of JSFR, such approaches are employed at the conceptual design stage.

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