Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
; Kato, Atsushi; Sato, Kazujiro
GENES4/ANP2003, 0 Pages, 2003/00
None
Negishi, Hitoshi; Sagayama, Yutaka; Sato, Kazujiro; ; Nakai, Ryodai; Ieda, Yoshiaki
Saikuru Kiko Giho, (20), 0 Pages, 2003/00
None
Kato, Atsushi; Sato, Kazujiro;
JNC TN1400 2002-015, 62 Pages, 2002/11
JNC-TN1400-2002-015.pdf:2.37MB
no abstracts in English
Negishi, Hitoshi; ; ; ; Sato, Kazujiro; Ieda, Yoshiaki
Saikuru Kiko Giho, (16), p.1 - 12, 2002/09
None
Sato, Kazujiro; Takano, Hideki*
JNC TY1410 2001-001, 57 Pages, 2001/04
JNC-TY1410-2001-001.pdf:2.69MB
no abstracts in English
; ; Takahashi, Nobutomo; Sato, Kazujiro;
Journal of Nuclear Science and Technology, 36(6), p.522 - 534, 1999/00
Times Cited Count：8 Percentile：53.67(Nuclear Science & Technology)None
Kamide, Hideki; ; ; Sato, Kazujiro; Hayashi, Kenji
Proceedings of 3rd International Conference on Nuclear Engineering (ICONE-3), 0 Pages, 1995/00
None
*; Maekawa, I.*; Sato, Kazujiro*
PNC TN9410 87-056, 139 Pages, 1987/03
The mass transport version (ver.MT) of three-dimensional thermal-hydraulic analysis code, COMMIX-PNC, has been developed to evaluate gas blow -down effects in the annulus between the MONJU closure head and plug port. The ver.MT has been validated through the analysis of the fundamental experiment of KCl transport and gas blow-down mock-up experiment. The fundamental experiments were carried out using a water cavity with 500 mm 
500 mm 50 mm in size. The experiments began pouring KCl solution into the inlet of the cavity. The calculated histories of the KCl concentration transient agreed well with the experiment. For the mock-up gas blow-down experiment, three gas flowrate cases, 0.05m
/min, 0.02m/min and 0.1m/min were calculated. Noble gas were predicted to reach the top part of the annulus only in the case with 0.1m/min flowrate. Through the application of the code to the MONJU configuration, the follwing have been effects of gas blow-down as obtained : [Normal gas blow condition] F.P. gases didn't enter into the annulus. [Half gas blow condition] The gases with 10
% concentration reached the location of 3905 mm above the bottom of shielding plug, and [Gas blow trip condition] The gases with 13% concentration reached the door valve in 1000 sec. simulation. From the above results and their consistency with the evaluation of gas blow-down effects by the experimental correlation derived from the gas blow-down experiments, the correlation can be applicable to a complicated annulus like that of the MONJU.
Sato, Kazujiro; ; Kamide, Hideki; Murakami, Takanori;
Donen Giho, (60), p.53 - 57, 1986/12
None
*; Hayashi, Kenji; *; Sato, Kazujiro*
PNC TN941 85-152, 48 Pages, 1985/10
A simple geometry model test was performed to study natural circulation thermohydraulics in the Monju EX-Vessel Storage Tank (EVST), using water as a working fluid. Thermohydraulic data in steady state natural circulation were obtained from measurements of temperature, velocity and flow pattern profiles. Also, analytical results by COMMIX-DRACS were compared with experiments to examine applicability of the code to the buoyancy dominated flow. The flow in the model is extremly complex and varies in space and time, however, it is comfirmed that temperature variations in circumferential directions are nearly uniform. COMMIX-DRACS results indicate relatively good agreement with the basic physical effects shown by the experimental data. Both analytical flow and temperature profiles match with experiments. Thus, the COMMIX-DRACS code is capable of predicting thermohydraulics in the Monju EVST by the use of the proposed analytical model in the present paper.
Hayashi, Kenji; Matsumoto, Akira; Sato, Kazujiro
PNC TN941 85-121, 24 Pages, 1985/08
Sodium vapor (or mist) concentration in argon cover gas space was measured at relatively low temperature conditions. The tests was conducted using a small pot in which natural convection was inducted by the axial temperature difference. Sodium temperature 150
300
C and axial temperature difference in the pot are dependent variables in the tests. Sodium concentration in the cover gas space above the surface of relatively low temperature (150 300C) logarithmically decreases with decreasing sodium surface temperature. The present data exist in the extrapolation zone of the previous data obtained in the high temperature range, however, there is almost two orders of magnitude difference between the present data (210
9 10 g/cm) and the saturated vapor concentration (1 10
g/cm) at 200C. The cause of this large difference is considered to be attributed to mist formation at just above sodium surface where sodium vapor should nucleate due to steep temperature drop. The data obtained in the present tests will be reflected on the evaluation of sodium deposition rates for the low temperature annulus wall of the Monju rotating shield plug.
*; *; Sato, Kazujiro*
PNC TN941 85-98, 51 Pages, 1985/07
Hydraulic tests of the corse control rod assembly were performed to collect experimental data required for an evaluation of core flow distribution in the Monju reactor by using the Water Test Loop. For simulating reactor operating conditions, pressure loss and intra-flow distribution characteristics were measured with varying an axial location of the control rod bundle, which is movable in the range from 0 mm to 1000 mm. In addition, effects of the FIV (Flow-Induced Vibration) restraint buttons were examined by mounting two accelerometers at the lower end of the bundle protect tube. Pressure loss characteristics are independent of axial locations of the control rod bundle in the range from 300 mm to 900 mm. Intra-flow distributions are also constant above 200 mm and about 45% of the total flowrate flows into the control rod bundle. The axial location of the bundle varies from 465 mm to 765 mm during the first and last core operations of Monju. Thus, it is indicated that hydraulic charaeteristics of the control rod assembly can be treated as to be constant in working ranges. Amplitudes of FIV at the measurement points do not decrease with mounting the FIV restraint buttons. However, the buttons are effective in decreasing anisotropic rolling of the protect tube. Experimental data obtained in the present tests will be reflected on the control rod assembly and core thermo-hydraulic designs of Monju.
*; *; Sato, Kazujiro*
PNC TN945 85-07, 85 Pages, 1985/06
The 1/2-scale model of the LMFBR Monju reactor was constructed to study core flow distribution using water as a working fluid. This paper presents the preliminary test results for core elements (core subassemblies, control rods, etc.) in the model concerning with hydraulic characteristics. Calibrations of the instrumental turbine flow-meters in the subassemblies have been also performed in the test. Hydraulic characteristics of three core elements in the each flowrate zone were measured and then experimental equations of flow resistance coefficients, 
, were obtained as a function of Reynolds Number, Re. Also, correlations between the pulse count number and the flowrate were obtained in the calibration tests. The above basic data about the 1/2-scale reactor model are essential for both the experiment and the analysis of the follow-on core flow distribution test. The results will be reflected on the core thermo-hydraulic design of Monju through the follow-on test.
Miyaguchi, Kimihide*; Sato, Kazujiro; Iguchi, Tatsuro*
PNC TG033 82-01(5), 16 Pages, 1982/01
Several series of thermal and hydraulic experiments have been conducted using full mock-ups of JOYO fuel subassemblies. This paper describes the test results of cross-flow mixing and pressure loss in wire-wrapped pin bundles. It also presents the comparison between cross-flow mixing experiments and calculations by SWIRL code.
H.Araki*; T.Komab*; Sato, Kazujiro*; Y.Ozaki*; H.Inuzi*; K.Koday*; Y.Shino*
PNC TN941 81-114, 11 Pages, 1981/06
Some studies for confirmation of the usefulness of the measurements of boiling sound and fluctuations of sodium temperature, flowrate and neutron flux for detection of sodium boiling have been curried out in Japan. For the development of acoustic detection system, acoustic transmission decay and velocity change by fuel pins, gas plenum and neutron shield were measured. Acoustic transmission from the reactor core to the detectors which were installed in the upper core structure and the upper flange of reactor vessel were studied theoretically and experimentally using the reactor vessel mockup. Sodium boiling acoustic spectra which were dependent of detector positions were obtained in the out-of-pile experimental facility. Acoustic background caused by flow and electric noise were measured in the Experimental Fast Reactor. On the other hand, coolant temperature and flowrate fluctuations caused by the sodium boiling were measured by the Temp/Flow detector which were mounted on the outlet of fuel subassembly. Simulation tests by computer on the correlation method between neutronic and acoustic signals were carried out. Current status and future plans about these studies in Japan are summarized briefly in this paper.
Fujimoto, Tetsuro*; Sato, Kazujiro*; Takahashi, June*
PNC TN951 76-14, 59 Pages, 1976/09
The 4,800-hours flow test for MONJU dummy core fuel subassemblies was carried out at sodium flow rate of 19.5 - 20.2 kg/sec. subassembly, sodium temperature of 600
, oxygen impurities in sodium of 2 - 2.5ppm which are about the same condition as that for MONJU core fuel subassemblies. The result of this test showed that the pressure loss of dummy core fuel subassemblies increased up to 11% in 3,000 hours and became constant after that. The integrity of the dummy subassemblies is planned to be investigated by disassembling them and making a material examination. On the other hand,the pressure loss increase for these subassembkies was estimated at 5.4% in case of water flow test which was carried out before and after spdium test. The cause of this difference should be investigated hereafter.
