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Featured researches published by Toshikuni Isozaki.


Nuclear Engineering and Design | 1994

Results of reliability test program on light water reactor piping

Katsuyuki Shibata; Toshikuni Isozaki; S. Ueda; Ryoichi Kurihara; Kunio Onizawa; A. Kohsaka

Abstract The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event. In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping. In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.


Nuclear Engineering and Design | 2000

Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP under PTS

Yinbiao He; Toshikuni Isozaki

One of the most severe accident conditions of a reactor pressure vessel (RPV) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at a high level. Such an accident is called pressurized thermal shock (PTS) transient according to the 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w=0.05∼0.9 were used to probe what kind of flaw and what kind of transient are most detrimental for the RPV in the end of life (EOL). Both the linear elastic and elastic–plastic material models were used in the stress analysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were investigated for different PTS transients. The fracture evaluation for the RPV in question under PTS transients, comparing with the material crack initiation toughness KIC, is performed in this paper.


Nuclear Engineering and Design | 1983

Transient analysis of blowdown thrust force under PWR LOCA

Toshikazu Yano; Noriyuki Miyazaki; Toshikuni Isozaki

Abstract The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces were obtained by Navier-Stokes momentum equation for a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a crifical flow condition was satisfied. The following results are obtained: 1. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena. 2. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08. 3. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one. 4. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break.


International Journal of Pressure Vessels and Piping | 1984

Experimental and analytical studies of four-inch pipe whip tests under PWR LOCA conditions

Noriyuki Miyazaki; Shuzo Ueda; Toshikuni Isozaki; Ryoichi Kurihara; Toshikazu Yano; R. Kato; Shyohachiro Miyazono

Abstract This paper presents experimental and analytical results of pipe whip tests performed under PWR LOCA conditions using a test pipe of 4-inch diameter and U-shaped restraints. In the tests, the effects of the overhang length on the pipe whip behavior of the pipe-restraints system were studied by measuring the strains and deformations of the test pipe and restraints, and the restraints forces. The equation for predicting the maximum strain at the outer surface of the pipe was derived using a static equilibrium condition. The calculated maximum strains at the outer surface of the pipe agree fairly well with experimental data. The dynamic response analysis of the pipe-restraints system was conducted by the finite element program ADINA. The applicability of the ADINA program to the pipe whip analysis is made clear through this analysis.


International Journal of Pressure Vessels and Piping | 1990

Measurement of leak-rate through fatigue-cracks in pipes under four-point bending and BWR conditions

Toshikuni Isozaki; Katsuyuki Shibata; H. Shinokawa; Shohachiro Miyazono

Abstract Leak-rate tests were performed using 114 mm and 165 mm (4 and 6 in) diameter, schedule 80 pipes made of austenitic stainless steel SUS304 and carbon steel STS42. Each pipe contained a through-wall fatigue crack and was mounted on a four-point bending machine of 400 kN maximum loading. Tests were done under a pressure of 7 MPa, with a subcooling temperature. The leak rate was measured by a Venturi flow meter and a differential pressure transducer attached to the pressure vessel. Comparisons of the effect of pipe material, diameter and crack angle were made. This paper shows that from a Leak-Before-Break viewpoint, the stainless-steel pipe is superior to the carbon-steel one, and that the pipe with the larger diameter is better than the one with the smaller diameter. No unstable fracture was observed in the tests.


Nuclear Engineering and Design | 1986

Experimental study of jet discharging test results under BWR and PWR loss of coolant accident conditions

Toshikuni Isozaki; Shohachiro Miyazono

Abstract This report describes the results of the jet discharging experiments conducted at the Japan Atomic Energy Research Institute. The tests were done under BWR and PWR Loss of Coolant Accident conditions using 4 inch, 6 inch and 8 inch test pipes, and varying distance between the pipe exit and the target plate. Simple and practical experimental formulae to estimate the maximum pressure on the target plate and maximum pressure distribution are given. Further, relations between pipe reaction thrust forces and jet impingement forces are described.


Nuclear Engineering and Design | 1984

TEST RESULTS OF JET DISCHARGE FROM A 4 INCH PIPE UNDER BWR LOCA CONDITIONS

Toshikuni Isozaki; Toshikazu Yano; Noriyuki Miyazaki; R. Kato; Ryoichi Kurihara; Shuzo Ueda; Shohachiro Miyazono

This report describes the result of tests on jet discharge of saturated water through a 4 inch discharge pipe from a pressure vessel having 4 m3 of inner volume, where the initial pressure was 69 kg/cm2 g and the temperature was 283°C. Behind the exit, a 10 ton load cell was installed to measure thrust force due to sudden jet discharge. In front of the pipe exit a 1000 mm diameter target disk was installed. The distance from the pipe exit to the target was 500 mm. Thirteen pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. The following conclusions were reached. 1. (1) Pipe reaction thrust is calculated using a homogeneous flow model at the early stage of blowdown where stagnation quality is low. It agrees closely with the experimental data. 2. (2) Maximum pressure on the target due to jet impingement is 1 kg/cm2 g when the distance from pipe exit to target is 500 mm. 3. (3) Mass flow rate was calculated by RELAP4/MOD5. Agreement between computed and experimental results was reached based on the assumption that CD = 0.63 for a 4 inch discharge pipe.


Nuclear Engineering and Design | 1983

Experimental studies of 4-inch pipe whip test under BWR loca conditions

Ryoichi Kurihara; S. Ueda; Toshikuni Isozaki; Noriyuki Miyazaki; Toshikazu Yano; R. Kato; Shohachiro Miyazono

Abstract Pipe whip tests or jet discharge tests have been performed at the Japan Atomic Energy Research Institute, which simulate the instantaneous circumferential guillotine break of primary coolant piping in nuclear power plants. The present paper describes the results of the pipe whip tests using test pipes of 4 inch diameter, under the BWR LOCA conditions, which were performed from 1979 to 1981. The tests were carried out at an initial pressure of about 6.8 MPa and an initial temperature of about 285°C. The test pipe was 114.3 mm (4 in) in diameter, 8.6 mm in thickness and 4500 mm in length. The four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from Type 304 stainless steel. The experimental parameters were the clearance (30, 50 and 100 mm) and the overhang length (250, 400, 550 and 1000 mm). The main purpose of these tests is to investigate the effects of the clearance and the overhang length on the pipe whip behavior. It has been clarified from the test results that a smaller clearance and a shorter overhang length causes the deformation of the pipe and restraints to be minimized, and the test pipe collapses near the setting point of the restraints with the overhang length of 1000 mm.


Nuclear Science and Engineering | 1984

An experimental study of blowdown thrust and jet forces for a pipe under boiling water reactor loss-of-coolant accident conditions

Toshikazu Yano; Toshikuni Isozaki; S. Ueda; Noriyuki Miyazaki; Ryoichi Kurihara; R. Kato; Shohachiro Miyazono

Blowdown thrust and jet impingement forces are simultaneously examined in jet discharge tests relating to an instantaneous pipe rupture accident. Tests were performed with a 6-in. pipe under boiling water reactor loss of coolant accident conditions. The initial pressure of the hot saturated water was 6.86 MPa. The time history of the blowdown thrust force just after the break, the jet thrust parameter of the pipe, the jet impingement force, the pressure and temperature distributions of the impinging jet, and the relationship between the thermal-hydraulic quantities and the thrust forces are examined.


Nuclear Engineering and Design | 1981

PRTHRUST-J1 code for calculation of blowdown thrust force and its experimental verification☆

Noriyuki Miyazaki; Ryoichi Kurihara; R. Kato; Toshikuni Isozaki; S. Ueda

Abstract This paper presents an outline of the PRTHRUST-J1 code for calculating blowdown thrust force and gives two numerical examples to show the effectiveness of this code. One numerical example is the problem of saturated steam blowdown. The blowdown thrust forces obtained from the PRTHRUST-J1 code were compared with those of the simplified method of Moody. Fairly good agreement was found between these two results. The other numerical example is the problem of jet discharging tests with stop valve performed in Japan Atomic Energy Research Institute. Analysis was carried out by varying the discharge coefficient. The analytical blowdown thrust force and pressure in the discharging nozzle were compared with experimental results. Qualitative agreement was found between the analytical and experimental results of the blowdown thrust force. Generally speaking, the blowdown thrust forces obtained from the experiment were between the analytical results for discharge coefficients of 1.0 and 0.6.

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Ryoichi Kurihara

Japan Atomic Energy Research Institute

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Shohachiro Miyazono

Japan Atomic Energy Research Institute

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R. Kato

Japan Atomic Energy Research Institute

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S. Ueda

Japan Atomic Energy Research Institute

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Toshikazu Yano

Japan Atomic Energy Research Institute

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Katsuyuki Shibata

Japan Atomic Energy Research Institute

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Shuzo Ueda

Japan Atomic Energy Research Institute

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Kunihisa Soda

Japan Atomic Energy Research Institute

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Kunio Onizawa

Japan Atomic Energy Research Institute

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