The Experts below are selected from a list of 360 Experts worldwide ranked by ideXlab platform

Hideki Kamide - One of the best experts on this subject based on the ideXlab platform.

  • development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled Fast Reactor
    Journal of Nuclear Science and Technology, 2015
    Co-Authors: Osamu Watanabe, Kazuhiro Oyama, Junji Endo, Norihiro Doda, Hideki Kamide, Takahiro Murakami, Yuzuru Eguchi
    Abstract:

    A natural circulation evaluation methodology has been developed to ensure the safety of a Sodium-Cooled Fast Reactor (SFR) of 1500 MW adopting the natural circulation decay heat removal system (NC-DHRS). The methodology consists of a one-dimensional safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a three-dimensional fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method for the hot spot temperature in the core. The safety analysis method and the three-dimensional analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a sodium test simulating a part of the primary system and the DHRS with about a 1/7 scale, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the three-dimensional analys...

  • selector valve failed fuel detection and location system for japan sodium cooled Fast Reactor
    Nuclear Technology, 2015
    Co-Authors: Kosuke Aizawa, Hideki Kamide, Kaoru Fujita, Naoto Kasahara
    Abstract:

    The Japan Sodium-Cooled Fast Reactor (JSFR) is studied as an advanced loop-type Sodium-Cooled Reactor. A selector-valve (SV) mechanism is adopted in the design of JSFR for its failed fuel detection...

  • development of flow induced vibration evaluation methodology for large diameter piping with elbow in japan sodium cooled Fast Reactor
    Nuclear Engineering and Design, 2011
    Co-Authors: Hidemasa Yamano, Hideki Kamide, Nobuyuki Kimura, Hiroyuki Ohshima, Masaaki Tanaka, Osamu Watanabe
    Abstract:

    Abstract This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan Sodium-Cooled Fast Reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the Reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.

  • gas entrainment allowance level at free surface and gas dynamic behavior of sodium cooled Fast Reactor
    Nuclear Engineering and Design, 2011
    Co-Authors: Akira Yamaguchi, Hideki Kamide, Takashi Takata, Hiroyuki Ohshima, Eisaku Tatsumi, Kei Ito, Jun Sakakibara
    Abstract:

    In a Sodium-Cooled Fast Reactor (SFR), inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The sources of the gas bubbles are entrainment and dissolution of the Reactor cover gas (argon) at the vessel free surface and emission of the helium gas that is produced as a result of disintegration of B4C control rod material. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, it is a key issue from the design and safety viewpoint and the allowance level is necessary regarding the gas entrainment at the free surface and the gas bubble concentration in the primary system. In the present study, a gas entrainment allowance level at the free surface is discussed and rationalized for the Japanese SFR (JSFR) design. The influence of the gas entrainment is evaluated using the void fraction at the core inlet. Design criteria for the acceptable level of the gas entrainment and gas concentration are proposed in consideration of the background level of gasses in the coolant. For the purpose, a plant dynamics code VIBUL has been developed to apply to the JSFR design to evaluate the concentration distribution of the dissolved gas and the free gas bubble in the JSFR system. Using the plant dynamics code for the bubble behavior, the background level of the free gas (void fraction at the core inlet) has been obtained. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment rate and the entrained bubble radius. Furthermore, the possibility of bubble removal and design requirement of the device is investigated to satisfy the allowance level. It is noted that the background level is already very low in comparison with the induced void reactivity by the void passing the Reactor core.

  • experimental study on thermal stratification in a Reactor vessel of innovative sodium cooled Fast Reactor mitigation approach of temperature gradient across stratification interface
    Journal of Nuclear Science and Technology, 2010
    Co-Authors: Nobuyuki Kimura, Hiroyuki Miyakoshi, Hideki Kamide
    Abstract:

    An innovative Sodium-Cooled Fast Reactor has been investigated as part of the Fast Reactor cycle technology development project (FaCT). Thermal stratification after a scram is one of the main thermal loads of a Reactor vessel (R/V). R/V has an upper inner structure (UIS), which consists of perforated horizontal plates and control rod guide tubes, and has a slit in the radial direction for fuel handling. The UIS slit causes an asymmetric flow pattern in R/V. A water experiment using a 1/10-scale model was carried out. A steep temperature gradient across the stratification interface was observed at the neighborhood of the UIS slit in the experiment. This means that the jet through the UIS slit entrains the bottom of the stratification interface. In order to mitigate the temperature gradient across the stratification interface, the height of a plug, which was installed in the upper plenum to infill a hole at the dipped plates where a fuel handling machine was inserted during a fuel exchange operation, was changed as the parameter in the experiment. The experimental result shows that the temperature gradient near the R/V wall in the case of the lower plug position was 21% smaller than that in the case of the higher plug position. This reduction of the temperature gradient was sufficient to maintain the structural integrity of the R/V wall against the thermal stratification.

Sahoe Lim - One of the best experts on this subject based on the ideXlab platform.

  • development of ultrasonic waveguide sensor for under sodium inspection in a sodium cooled Fast Reactor
    Ndt & E International, 2011
    Co-Authors: Youngsang Joo, Chang Gyu Park, Jaehan Lee, Jongbum Kim, Sahoe Lim
    Abstract:

    Abstract The Reactor core and in-vessel structures of a Sodium-Cooled Fast Reactor (SFR) cannot be examined visually due to the opaque sodium. The examination of the in-vessel structures is possible using ultrasonics to penetrate the sodium. A plate-type ultrasonic waveguide sensor using a leaky Lamb wave (A0 mode) has been developed for under-sodium visual inspection of the Reactor core and in-vessel structures. In the plate waveguide sensor, the A0 leaky Lamb wave is utilized for the single mode generation and the effective radiation capability in a fluid. The liquid wedge is applied for the generation of the A0 mode in the low frequency range. The long pulse tone-burst excitation should be applied to minimize the dispersion effect in 10 m long distance propagation of the A0 Lamb wave. And a novel technique which is capable of steering a radiation beam of a waveguide sensor without a mechanical movement of the waveguide sensor has been suggested. The control of the beam angle can be achieved by a frequency tuning method of the excitation pulse in the dispersive low frequency range of the A0 Lamb wave. A 10 m long ultrasonic waveguide sensor module which consists of a plate waveguide, a liquid wedge, an ultrasonic sensor, and an acoustical shielding protection tube has been designed and manufactured. The possibility of applying the ultrasonic waveguide sensor module to an under-sodium visual inspection has been investigated. The experimental tests such as the long distance propagation test of A0 Lamb wave, the beam profile measurements, and C-scanning experiments in water have been carried out for the performance of the ultrasonic waveguide sensor. The feasibility of the ultrasonic waveguide sensor technique has been successfully demonstrated.

Shoji Kotake - One of the best experts on this subject based on the ideXlab platform.

  • seismic isolation design for jsfr
    Journal of Nuclear Science and Technology, 2011
    Co-Authors: Shigeki Okamura, Kazuo Negishi, Seiji Kitamura, Yoshio Kamishima, Y. Sakamoto, Shoji Kotake
    Abstract:

    This paper describes the seismic design of Japan Sodium-Cooled Fast Reactor (JSFR), which includes the seismic condition, the seismic isolation system, and the seismic evaluation of the primary components. Since the design seismic loading is set out severely than ever since The Niigata-ken Chuetsu-oki Earthquake in 2007, an advanced seismic isolation system is aimed to reduce the seismic force loaded on the primary components of JSFR to be less than that of the previous seismic isolation system. The advanced seismic isolation system is developed by optimizing the performance based on the previous seismic isolation system considering the natural frequency of the primary components. The laminated rubber bearings thicker than the previous ones and oil dampers are adopted for the advanced seismic isolation system of SFR. The seismic evaluation of nuclear Reactor components applying the advanced seismic isolation system is performed and its feasibility is confirmed.

  • design study and r d progress on japan sodium cooled Fast Reactor
    Journal of Nuclear Science and Technology, 2011
    Co-Authors: Kazumi Aoto, Y. Sakamoto, Nariaki Uto, Mikio Toda, Takaya Ito, Shoji Kotake
    Abstract:

    This paper describes the progress of the design study and research and development (R&D) for the Japan Sodium-Cooled Fast Reactor (JSFR) implemented in the “Fast Reactor Cycle Technology Development (FaCT)” project. A Sodium-Cooled Fast Reactor with an electric power of 1,500MWe is targeted for commercialization at around 2050, and a demonstration Reactor assuming a power output from 500 to 750MWe is planned to start operation at around 2025. R&D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact Reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened highchromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to...

  • safety strategy of jsfr eliminating severe recriticality events and establishing in vessel retention in the core disruptive accident
    Journal of Nuclear Science and Technology, 2011
    Co-Authors: Ikken Sato, Shoji Kotake, Shigenobu Kubo, Yoshiharu Tobita, Kenji Kamiyama, Ryodai Nakai, Kazuya Koyama, Kensuke Konishi, Junichi Toyooka, Yuri S Vassiliev
    Abstract:

    In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the Reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropri...

  • development of advanced loop type Fast Reactor in japan
    Nuclear Technology, 2010
    Co-Authors: Shoji Kotake, Yoshio Kamishima, Y. Sakamoto, Kazumi Aoto, Nariaki Uto, Takatsugu Mihara, Shigenobu Kubo, Mikio Toda
    Abstract:

    This paper describes the current status of the design study and the related research and development for an advanced loop-type Fast Reactor: the Japan Sodium-Cooled Fast Reactor (JSFR). First, the development targets and the major design requirements are established. Then, a cost-down approach in which developing innovative technologies is key to being competitive with another future energy source is discussed. Here, the development status of several innovative technologies such as a two-loop cooling system, reliable Reactor system, simplified fuel-handling system, passive Reactor shutdown system, mitigation measure against a core disruptive accident, and minor actinide-bearing oxide fuel core is described. Last, a review of JSFR development and the demonstration plan for the innovative technologies are discussed.

  • a next generation sodium cooled Fast Reactor concept and its r d program
    Nuclear Engineering and Technology, 2007
    Co-Authors: Masakazu Ichimiya, Tomoyasu Mizuno, Shoji Kotake
    Abstract:

    Critical issues in the development targets for the future Fast Reactor(FR) cycle system, including Sodium-Cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of Sodium-Cooled Fast Reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercialization is described, to be followed up in a new framework of the Fast Reactor Cycle Technology development(FaCT) Project launched in 2006.

Greg Kelsall - One of the best experts on this subject based on the ideXlab platform.

  • thermodynamic analysis and preliminary design of closed brayton cycle using nitrogen as working fluid and coupled to small modular sodium cooled Fast Reactor sm sfr
    Applied Energy, 2017
    Co-Authors: Olumide Olumayegun, Meihong Wang, Greg Kelsall
    Abstract:

    Abstract Sodium-Cooled Fast Reactor (SFR) is considered the most promising of the Generation IV Reactors for their near-term demonstration of power generation. Small modular SFRs (SM-SFRs) have less investment risk, can be deployed more quickly, are easier to operate and are more flexible in comparison to large nuclear Reactor. Currently, SFRs use the proven Rankine steam cycle as the power conversion system. However, a key challenge is to prevent dangerous sodium-water reaction that could happen in SFR coupled to steam cycle. Nitrogen gas is inert and does not react with sodium. Hence, intercooled closed Brayton cycle (CBC) using nitrogen as working fluid and with a single shaft configuration has been one common power conversion system option for possible near-term demonstration of SFR. In this work, a new two shaft nitrogen CBC with parallel turbines was proposed to further simplify the design of the turbomachinery and reduce turbomachinery size without compromising the cycle efficiency. Furthermore, thermodynamic performance analysis and preliminary design of components were carried out in comparison with a reference single shaft nitrogen cycle. Mathematical models in Matlab were developed for steady state thermodynamic analysis of the cycles and for preliminary design of the heat exchangers, turbines and compressors. Studies were performed to investigate the impact of the recuperator minimum terminal temperature difference (TTD) on the overall cycle efficiency and recuperator size. The effect of turbomachinery efficiencies on the overall cycle efficiency was examined. The results showed that the cycle efficiency of the proposed configuration was comparable to the 39.44% efficiency of the reference cycle. In addition, the study indicated that the new configuration has the potential to simplify the design of turbomachinery, reduce the size of turbomachinery and provide opportunity for improving the efficiency of the turbomachinery. The findings so far revealed that the proposed two-shaft CBC with nitrogen as working fluid could be a promising power conversion system for SM-SFRs near-term demonstration.

Nariaki Uto - One of the best experts on this subject based on the ideXlab platform.

  • A summary of Sodium-Cooled Fast Reactor development
    Progress in Nuclear Energy, 2014
    Co-Authors: Kazumi Aoto, Robert Hill, Philippe Dufour, Yang Hongyi, Jean Paul Glatz, Yeong-il Kim, Yury Ashurko, Nariaki Uto
    Abstract:

    Abstract Much of the basic technology for the Sodium-Cooled Fast Reactor (SFR) has been established through long term development experience with former Fast Reactor programs, and is being confirmed by the Phenix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicated to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.

  • design study and r d progress on japan sodium cooled Fast Reactor
    Journal of Nuclear Science and Technology, 2011
    Co-Authors: Kazumi Aoto, Y. Sakamoto, Nariaki Uto, Mikio Toda, Takaya Ito, Shoji Kotake
    Abstract:

    This paper describes the progress of the design study and research and development (R&D) for the Japan Sodium-Cooled Fast Reactor (JSFR) implemented in the “Fast Reactor Cycle Technology Development (FaCT)” project. A Sodium-Cooled Fast Reactor with an electric power of 1,500MWe is targeted for commercialization at around 2050, and a demonstration Reactor assuming a power output from 500 to 750MWe is planned to start operation at around 2025. R&D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact Reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened highchromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to...

  • development of advanced loop type Fast Reactor in japan
    Nuclear Technology, 2010
    Co-Authors: Shoji Kotake, Yoshio Kamishima, Y. Sakamoto, Kazumi Aoto, Nariaki Uto, Takatsugu Mihara, Shigenobu Kubo, Mikio Toda
    Abstract:

    This paper describes the current status of the design study and the related research and development for an advanced loop-type Fast Reactor: the Japan Sodium-Cooled Fast Reactor (JSFR). First, the development targets and the major design requirements are established. Then, a cost-down approach in which developing innovative technologies is key to being competitive with another future energy source is discussed. Here, the development status of several innovative technologies such as a two-loop cooling system, reliable Reactor system, simplified fuel-handling system, passive Reactor shutdown system, mitigation measure against a core disruptive accident, and minor actinide-bearing oxide fuel core is described. Last, a review of JSFR development and the demonstration plan for the innovative technologies are discussed.