Shoichi Moriya
Central Research Institute of Electric Power Industry
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Nuclear Engineering and Design | 1994
Y. Eguchi; Kazuhiko Yamamoto; T. Funada; Nobukazu Tanaka; Shoichi Moriya; Kouichi Tanimoto; K. Ogura; Takuro Suzuki; Isamu Maekawa
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 110-th, 16-th, 13-rd and11.6-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.
Nuclear Engineering and Design | 1990
Nobukazu Tanaka; Shoichi Moriya; Satoru Ushijima; Tomonari Koga; Yuzuru Eguchi
Abstract The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.
Nuclear Engineering and Design | 1990
Shoichi Moriya; Iwao Ohshima
Abstract The effects of working fluids on temperature fluctuation phenomena have been studied, with special reference to similarity rules for the thermal striping phenomena in LMFBRs. Temperature measurements were made in non-isothermal coaxial jets using sodium, water and air as working fluids. These experimental results indicated the dependence of the characteristics of average and root-mean-square values for the temperature and peak-to-peak amplitudes and periods of temperature wave-form signals on Reynolds and Peclet numbers. It may be concluded that the characteristics of temperature fluctuation due to turbulent mixing in LMFBRs can be evaluated from the results of model tests using water and/or air if the Reynolds and Peclet numbers are sufficiently large.
Nuclear Engineering and Design | 1987
Shoichi Moriya; Nobukazu Tanaka; N Katano; A Wada
Thermal transient water tests were performed using a simplified hot plenum model of a pool-type LMFBR to examine the fundamental characteristics of thermal stratification during reactor trips. These tests were divided into two series to clarify the effects of (1) Reynolds number and (2) Richardson number on the behavior of thermal stratification. The results of the first series indicated that the characteristics of thermal stratification were independent of Reynolds number in the range of Re > 104. It was concluded from the results of the second series that the rising speed of the thermal interface and the temperature gradient at the interface varied proportionally with Ri−12 and Ri12, respectively. In addition, large scale internal waves, called internal standing waves, were observed under the condition of Re > 104 and 1 < Ri < 5.
Nuclear Engineering and Design | 1990
Satoru Ushijima; Nobukazu Tanaka; Shoichi Moriya
Abstract In the present study, non-isothermal coaxial jets are numerically investigated and the turbulence quantities calculated with a second order closure model are compared with the experimental results and the validity is examined.
10th International Conference on Nuclear Engineering, Volume 3 | 2002
Tsutomu Kawamura; Kouji Shiina; Masaya Ohtsuka; Isao Tanaka; Hiroshi Hirayama; Kouichi Tanimoto; Toshihiko Fukuda; Akihiro Sakashita; Jun Mizutani; Yasuhiko Minami; Shoichi Moriya; Haruki Madarame
Thermal striping tests in mixing tees with hot and cold water were conducted for three types of flow conjunctions in order to establish an evaluation method for high-cycle thermal fatigue of piping systems. Two kinds of examinations were planned. The preliminary tests were flow visualization tests carried out using acrylic pipes to obtain flow pattern characteristics and flow temperature fluctuations. The main tests were temperature fluctuation measurement tests carried out using metal pipes to evaluate the unsteady heat transfer coefficient based on measured temperature fluctuations of fluid and pipe wall. This paper reports visualization test results. The flow patterns were visualized by injection of methylene blue and compared with flow analysis results by the k-e turbulence model. Temperature fluctuations of fluid 3mm from the inner pipe wall were measured with C-A thermocouples. Fundamental features such as locations with a large fluctuating temperature, the fluctuating temperature amplitude and its frequency were identified.Copyright
Nuclear Science and Engineering | 1984
Nobukazu Tanaka; Shoichi Moriya; Akira Wada
The most important problem in the thermal-hydraulic designs of the pool-type fast breeder reactor is to estimate the thermal conditions affecting the vessel and/or internal structures during both steady and transient operations. The severity of these conditions in the Japanese pool-type reactor, which will be reinforced and equipped with special devices for seismic demands, is apt to be much greater than for other countries. Water tests and thermal-hydraulic analyses have been performed to study such conditions. The effects of the elevations of upper internal structure discharge and intermediate heat exchanger intakes on flow patterns, free surface disturbance, and thermal stratification in the hot plenum have been estimated. From the results of the experiments, suitable elevations could be recommended by comparing some thermal-hydraulic characteristics. The calculations agreed well with the experimental results for the steady-state flow patterns and thermal transients, with the exception of thermal stratification.
Nuclear Engineering and Design | 1987
A. Masui; T. Koga; Shoichi Moriya; Nobukazu Tanaka
Abstract Thermal hydraulic experiments and analyses in water have been performed to study in-vessel thermal hydraulic phenomena of a pool-type LMFBR. These phenomena are highly turbulent, and we need an in-vessel hydraulic analysis code with which the turbulent phenomena can be accurately estimated. We have developed and applied a new analysis code to which the two-equation turbulence ( k - σ ) model was introduced. This code was applied to the simulation of the above-mentioned water tests and we have evaluated its effectiveness by comparison with experimental results. The k - σ model can well simulate thermal hydraulic phenomena in general except corners, outlet and density stratified areas where nonisotropic features are dominant.
Transactions of the Japan Society of Mechanical Engineers. B | 2004
Yoshiyuki Kondo; Koichi Tanimoto; Tadashi Shiraishi; Shigeki Suzuki; Kenji Ogura; Kouji Shiina; Toshihiko Fukuda; Naoki Chigusa; Shoichi Moriya
Transactions of the Japan Society of Mechanical Engineers. B | 2005
Kouji Shiina; Tsutomu Kawamura; Masaya Ohtsuka; Tadashi Mizuno; Masakazu Hisatsune; Kenji Ogura; Kouichi Tanimoto; Toshihiko Fukuda; Yasuhiko Minami; Shoichi Moriya; Haruki Madarame