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Featured researches published by Isao Sumida.


Nuclear Technology | 1988

Conceptual Design and Thermal-Hydraulic Characteristics of Natural Circulation Boiling Water Reactors

Yoshiyuki Kataoka; Hiroaki Suzuki; Michio Murase; Isao Sumida; Tetsuo Horiuchi; Minoru Miki

A natural circulation boiling water reactor (BWR) with a rated capacity of 600 MW(electric) has been conceptually designed for small- and medium-sized light water reactors. The components and syste...


Nuclear Technology | 1992

Experiments on Convection Heat Transfer Along a Vertical Flat Plate Between Pools with Different Temperatures

Yoshiyuki Kataoka; Tohru Fukui; Shigeo Hatamiya; Toshitsugu Nakao; Masanori Naitoh; Isao Sumida

This paper reports that to evaluate the heat removal capability of an external water wall-type containment vessel, which is a passive system for containment cooling, thermal-hydraulic behavior in the suppression and outer pools has been examined experimentally. The following results are obtained: A thermal stratification boundary, which separates the pools into an upper high-temperature region and a lower low-temperature region, is observed just below the vent outlet. The natural-convection heat transfer coefficients (HTCs) for the downward and upward flows that appear inside and outside the primary containment vessel wall are measured. The condensation HTCs in the presence of non-condensable gas, which affect heat transfer between the wet well and the outer pool, are measured along the long wall. The capability for decay heat removal in the external water wall-type containment vessel for a 600-MW (electric) plant is evaluated based on these results and is found to be large enough.


Nuclear Technology | 1990

Conceptual design and safety characteristics of a natural-circulation boiling water reactor

Yoshiyuki Kataoka; Hiroaki Suzuki; Sigeo Hatamiya; Michio Murase; Isao Sumida; Tetsuo Horiuchi; Minoru Miki

The Hitachi simplified boiling water reactor (BWR) is a natural-circulation BWR with a rated capacity of 600 MW (electric). It has been designed for the classes of small- and medium-sized light water reactors. The power density is ∼70% of that in current BWRs because of natural circulation, and the reactor pressure vessel is larger. The components and systems have been simplified by eliminating pumped recirculation systems and pumped emergency core cooling systems. Consequently, the volume of the reactor building is ∼50% that of current BWRs, with the same rated capacity. The construction period is also shorter. In addition, the lower power density allows continuous operation for a 23-month period. The safety characteristics of this BWR are investigated during transient and accident conditions, and the high standards of its simple safety systems are shown


Nuclear Technology | 1978

Impurity Interaction Analysis in Mesh-Packed Cold Traps

Michio Murase; Isao Sumida; Koichi Kotani; Hajime Yamamoto

An impurity interaction code for a mesh-packed cold trap was developed, and the properties of the packed trap, such as the trapping efficiency and capacity, were calculated using this code. As a result, the trapping efficiency and capacity are greatly affected by the packing ratio and the wire diameter of the mesh packing. Precipitation of impurity is promoted at the entrance to the mesh packing. This facilitating effect hastens flow blockage and decreases the capacity. Moreover, the capacity decreases with increase in the efficiency.


Journal of Nuclear Science and Technology | 1978

Theory of Hydraulic Stability of Boiling Channels: Analysis of Structure of Problem

Isao Sumida; Toshio Kawai

A framework of boiling channel stability theory is analyzed. The fundamental equations are those of STABLE code: Three conservation laws of mass, energy and momentum applied to one-dimensional channel, together with Bankoff slip and Marinelli-Nelsons pressure drop correlation. These equations are analyzed to yield “Void Equation”, “Linearized Void Equation”, “Volume Conservation Law” and the “Flow Impedance” R(s), defined by the dynamic response of pressure drop to the inlet flow. The impedance contains all the information such a stability index, dominant frequency and damping ratio. It is shown that R is a sum of the form R IA+N F −1 R D+N R R R+N OR, where Ns are non-dimensional parameters and Rs characteristic impedances determined by three kinds of parameters, Nx , Ns and the power distribution parameter. Systematic edition of the characteristic impedances according to the non-dimensional parameters will reduce the need for case-by-case STABLE calculations. Hydraulic stability of BWR channels under...


Nuclear Engineering and Design | 1985

Dimensionless groups of blowdown experiments

Takashi Ikeda; Atsuo Yamanouchi; Isao Sumida; Masanori Naitoh

In order to clarify the limit of application of the dimensionless groups derived in a former study, numerical analyses were performed on the blowdown with different initial conditions. Dimensionless groups studied were π20 (fractional change rate of residual coolant mass), π30 (fractional change rate of energy due to heat generation), and π0 (= π30/π20). As the reference, the blowdown was considered from the pressure vessel initially filled with saturated water (volumetric fraction of vapor αi = 0) at 6.9 MPa. Numerical analyses based on the lumped model description of blowdown indicated that the dimensionless groups obtained were useful for the blowdowns with not only different volume and discharge area of the vessel, but also different initial pressures (6–8 MPa) and amounts of coolant (αi = 0-0.2). Clarified groups were applied to the comparison of the BWR recirculation line break tests data between two differently scaled facilities. The comparison yielded good agreement in overall trends of both test results and confirmed the validity of the dimensionless groups. Further, the dependence of the major event sequence on the break area during recirculation line break tests was generalized using the dimensionless groups.


Journal of Nuclear Science and Technology | 1976

Fundamental Experiment of Potassium Heat Exchanger Using Principle of Heat Pipe

Isao Sumida; Koichi Kotani

Abstract In order to provide compact and reliable sodium equipments including a steam generator, performance tests are conducted with a potassium heat exchanger, which is featured by the separate construction of primary and secondary coolant systems. A small amount of potassium plays a role as an intermediate media of heat transportation between these two coolant systems. Heat is transfered by evaporation and condensation of potassium on the surfaces of the primary and the secondary coolant pipings, respectively. The tests are performed in the temperature range of 200-300°C and the maximum heat transfer reaches 1.3 kW (heat transfer rate at the primary heating source: 8.6 W/cm2 at 300°C). The experimental results are analyzed by using Langmuirs and Schrages equations and close agreement between experiment and theory is obtained.


Archive | 1991

Reactor containment facilities

Kenji Tominaga; Tetsuo Horiuchi; Tsuyoshi Niino; Shouichiro Kinoshita; Shozo Yamanari; Masanori Naitoh; Tohru Fukui; Michio Murase; Yoshiyuki Kataoka; Masataka Hidaka; Isao Sumida


Archive | 1988

Method of and apparatus for controlling power of natural circulation reactor

Hiroaki Suzuki; Yoshiyuki Kataoka; Michio Murase; Kotaro Inoue; Isao Sumida; Shozo Yamanari; Masaki Matsumoto; Satoshi Miura; Koji Hashimoto


Archive | 1991

Nuclear reactor installation

Tadashi Fujii; Yoshiyuki Kataoka; Tohru Fukui; Masataka Hidaka; Toshitsugu Nakao; Shigeo Hatamiya; Hiroaki Suzuki; Masanori Naitoh; Isao Sumida; Kenji Tominaga; Tsuyoshi Niino

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