Demonstration Fast Breeder Reactor Research Articles

Safety approach of DFBR design study in Japan

The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area. The design concept of these safety features is described in this paper.

Development of elevated temperature structural design guide for Japanese demonstration breeder reactor

Programs to develop the “elevated temperature structural design guide for the demonstration fast breeder reactor” (DDS) in Japan have been conducted since 1987. The DDS is to be developed on the basis of the “elevated temperature structural design guide for class 1 components of prototype fast breeder reactors” (ETSDG), by considering structural and material features of the demonstration fast breeder reactor (DFBR) and incorporating results of the latest R&D. This paper describes the progress of the R& D concept of the DDS, and discusses some typical results of current studies on the DDS.

２６４． 高速増殖実証炉開発の現状と実用化見通し

The Demonstration Fast Breeder Reactor (DFBR) is the next step of FBR development following the prototype fast breeder reactor MONJU. The DFBR is now under development by The Japan Atomic Power Co. (JAPC) under the sponsorship of 9 Japanese electric power companies and Electric Power Development Co., Ltd. The JAPC has been performing the design study and R & D for DFBR in cooperation with Power Reactor and Nuclear Fuel Development Corp. (PNC), Central Research Institute of Electric Power Industry (CRIEPI) and Japan Atomic Energy Research Institute (JAERI). This report describes the prospect of FBR commercialization and the current status of new technology for DFBR and innovative technology toward FBR commercialization.

Development of the demonstration FBR and its commercialization

The FBR being developed in Japan is expected to operate commercially by the year 2030 or so. The construction of the first demonstration FBR “DFBR-1”, which follows the experimental FBR “JOYO” and the prototype FBR “MONJU” is scheduled for early 2000s. The Atomic Energy Commission adopted these policies in the new version of “The Long Term Program for Research, Development and Utilization of Nuclear Energy”. The Japan Atomic Power Company has progressed design studies of the DFBR-1 and R&Ds. Electric utilities settled the major design specifications of the DFBR-1 such as the top entry loop type as the reactor type and about 660 MWe of plant capacity. To achieve the goals for the commercialization of FBRs, such as a high level of safety and reliability comparable with a LWR and further improvement of economy, many innovative technologies and system improvements will be required in commercial FBRs. This paper describes the present status of development of the DFBR-1.

ＦＢＲ構造設計エキスパートシステムの開発

The characteristics of structural design processes for nuclear components can be summarized as follows :(1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert important roles in the design process. Taking these issues into consideration, a new expert dataflow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw. the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures.

Gas entrainment in the IHX vessel of top-entry loop-type LMFBR

The effects of scale and fluid properties on gas entrainment onset were experimentally studied for the free surface flow in an IHX vessel of a proposed Japanese demonstration FBR. Water experiments were performed using the 1 10 - th, 1 6 - th, 1 3 - rd and 1 1.6 - th scaled models of the IHX. The experimental results indicate that the scale has a strong effect on gas entrainment onset and that the critical Fr number decreases in proportion to the −0.5-th power of the scale. Applicability of a computational method to the gas entrainment is also discussed by comparing computational and experimental results.

Flow Characteristics Related to the Vibration Problem of Flexible Cylindrical Weir Due to Fluid Discharge.

The present study deals with the unsteady characteristics of flow concerning flow-induced vibration due to fluid discharge over a flexible cylindrical weir, which occurred in the French demonstration fast breeder reactor, Super Phenix-1. The theoretical analysis of sloshing in a double cylindrical vessel and the drop of the free surface at the fall-in section and the numerical analysis of the inner flow were carried out. In the numerical analysis, unsteady and 2D inner flow with periodic vortex formation and shedding were simulated using the boundary element method and the discreet vortex method. As a result, it was obtained that the frequency of simulated unsteady flow was in good agreement with the natural frequency of sloshing, and the value of frequency converted into the case of the 1/5-scale test model of the Japanese demonstration FBR was about 6 Hz, which is near the frequency of the fluid elastic mode of unstable vibrations.

Exploratory research on creep and fatigue properties of 9Cr-steels for the steam generator of an FBR

Research on the applicability of 9Cr-steels to the steam generator of the demonstration fast breeder reactor was performed by the Subcommittee of the Japan Welding Engineering Society as a four-year program from 1985. In this program, exploratory tests, which included tensile, creep rupture and low-cycle fatigue tests, were conducted on three kinds of 9Cr-steels (Mod.9Cr-1Mo, 9Cr-1Mo-V-Nb, and 9Cr-2Mo) and their weldments. This paper describes the summary of results obtained in this program. Among the tested 9Cr-steels, Mod.9Cr-1Mo steel shows the best creep rupture strength and its weldment indicates almost the same level of creep rupture strength and the base metal at 500 and 550°C. The low-cycle fatigue properties of Mod.9Cr-1Mo steel is also discussed from its relation to the tensile properties.

Fatigue life assessment in steel structures in the presence of initial defects: Nair, E.M.S., Moorthy, P.A.S., Ganapathy, C. and Nair, N.G. Conf. Fatigue and Fracture in Steel and Concrete Structures, ISFF '91, Madras, India, Dec 1991 Vol 2 (Oxford & IBH Publishing Co. Pvt. Ltd, New Delhi, India, 1991) 1035–1046

Research on the applicability of 9Cr-steels to the steam generator of the demonstration fast breeder reactor was performed by the Subcommittee of the Japan Welding Engineering Society as a four-year program from 1985. In this program, exploratory tests, which included tensile, creep rupture and low-cycle fatigue tests, were conducted on three kinds of 9Cr-steels (Mod.9Cr-1Mo, 9Cr-1Mo-V-Nb, and 9Cr-2Mo) and their weldments. This paper describes the summary of results obtained in this program. Among the tested 9Cr-steels, Mod.9Cr-1Mo steel shows the best creep rupture strength and its weldment indicates almost the same level of creep rupture strength and the base metal at 500 and 550°C. The low-cycle fatigue properties of Mod.9Cr-1Mo steel is also discussed from its relation to the tensile properties.

Safety Design Study of Fast Breeder Reactors in Japan

Two fast breeder reactor (FBR) concepts, the tank type and the loop type, have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle of the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant.

Development, operational experience and implications for future design of FBRS in Japan

Joyo, the 100 MW t experimental reactor, has been successfully operated since 1977, and Monju, the 280 MW e prototype FBR, is under construction, with the first criticality planned for 1992. To promote FBR R&D efficiently — including the demonstration FBR (DFBR) programme — a steering committee for R&D was organized in 1986 by the Japan Atomic Power Company, the Power Reactor and Nuclear Fuel Development Corporation, the Japan Atomic Energy Research Institute and the Central Research Institute of Electric Power Industry. A design study of the DFBR is now underway to define its basic specifications by 1990. R&D for Monju, DFBR and future commercial FBRS has been done (1) to improve key technologies developed through the Joyo and Monju programmes; (2) to develop innovative technologies to make FBRS commercial; (3) to promote FBR development in conjunction with the development of the FBR fuel cycle.

Fluid-elastic vibration of flexible overflow weir.

The present paper describes the fundamental feature of the fluid-elastic vibration of flexible overflow weir, as observed in the French demonstration fast breeder reactor, Super Phenix-1. In the experimental study, the instability criterion of the fluid-elastic vibration was investigated using a simple experimental apparatus of a rectangular tank separated by a flexible weir. A spring-mass model for sloshing motion of liquid contained in the tank was developed to clarify the mechanism of the instability. The instability condition was analytically derived from the equations of the spring-mass model. The equations of the spring-mass model were also computationally integrated in time to simulate the timewise evolution of the fluid-elastic vibration. The comparison between the experimental and theoretical results indicate that the present theoretical model is capable of predicting the general characteristics of the vibration observed in the experiment. The present study revealed that the lag-time of the water fall at the weir is the most influential parameter on the fluid-elastic instability of a flexible overflow weir.

Status of Axial Heterogeneous Liquid-Metal Fast Breeder Reactor Core Design Studies and Research and Development

The current status of axial heterogeneous core (AHC) design development in Japan, which consists of an AHC core design in a pool-type demonstration fast breeder reactor (DFBR) and research and development activities supporting AHC core design, is presented. The DFBR core design objectives developed by The Japan Atomic Power Company include (a) favorable core seismic response, (b) core compactness, (c) high availability, and (d) lower fuel cycle cost. The AHC concept was selected as a reference pool-type DFBR core because it met these objectives more suitably than the homogeneous core (HOC). The AHC core layouts were optimized emphasizing the reduction of the burnup reactivity swing, peak fast fluence, and power peaking. The key performance parameters resulting from the AHC, such as flat axial power/flux distribution, lower peak fast fluence, lower burnup reactivity swing, etc., were evaluated in comparison with the HOC. The critical experiments at the Japan Atomic Energy Research Institute’s Fast Critical Assembly facility demonstrate the key AHC performance characteristics. The large AHC engineering benchmark experiments using the zero-power plutonium reactor and the AHC fuel pin irradiation test program using the JOYO reactor are also presented.

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