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Yoshida, Eiichi; Hirakawa, Yasushi; Yatabe, Toshio
JAEA-Technology 2012-033, 177 Pages, 2012/11
JAEA-Technology-2012-033.pdf:17.98MB
In JAEA, lots of tests using sodium had been carried out on the development of sodium component systems and sodium technologies for the experimental reactor JOYO and prototype reactor MONJU. When research and development has come to the end of the first stage for these reactors, those sodium test facilities has dismantled and attached sodium has been cleaned. Lots of experiences and knowledge of sodium cleaning treatment technologies has accumulated. In order to use those experiences and knowledge effectively for future sodium technology and research on the next generation fast reactors, experiences and knowledge of sodium technologies for typical systems and components has been evaluated and knowledge related important topics has been rearranged. Based on those evaluation and rearrangement, technical guidelines of sodium cleaning treatment technologies has been proposed for the purpose of effective reference of the past experiences and knowledge.
Futagami, Satoshi; Kurihara, Akikazu; Yatabe, Toshio
JNC TN9400 2005-042, 82 Pages, 2005/08
JNC-TN9400-2005-042.pdf:10.42MB
As a basic test for establishing the mechanism-theoretical evaluation technique of the water leak phenomenon in steam generator, free jet test(Run-FJ series) was carried out with spouting the steam in the stagnant liquid-sodium without heat transfer tubes. In the test, the temperature distribution was measured in detail, and the effect of the cover gas pressure was evaluated. As a test parameter, 2 cases of the cover gas pressure were selected(Run-FJ-1:0.5 MPa (gauge), Run-FJ-2:0.05MPa (gauge)). The water leak rate was made to be intermediate scale leak rate (0.2 kg/sec). From this test, following conclusions were obtained on sodium-water reaction jet. (1) Phase condition of the reaction jet is estimated from temperature oscillation intensity distribution. In the region where gas phase seems to become dominant, it shows the high temperature oscillation, and it agrees with the tendency in the distribution of the void output. For this reason, the heat transfer coefficient would decrease in the high oscillation area. (2) High-temperature region of Run-FJ-1 and Run-FJ-2 were narrower than Run-HT-1(heat transfer tubes existed). In comparison Run-FJ-1 with Run-FJ-2, the width of high-temperature region was similar. However, the length of high-temperature region of Run-FJ-2 is longer than Run-FJ-1 (Run-FJ-1: about 25cm, Run-FJ-2: about 50 cm). (3) Temperature distribution of the reaction jet depends on the cover gas pressure. It is because the behavior of the reaction product, hydrogen, seems to become dominant in the downstream of the reaction jet, and the behavior of the flow condition in the upstream of the reaction jet is effected by the jet stream. (4) Maximum temperature of the reaction jet is effected by the cover gas pressure (Run-FJ-1: 1048 degree, Run-FJ-2: 979 degree). It is considered that the boiling point of liquid sodium influences the maximum temperature, and that change of the gas phase by the cover gas pressure also influences the maximum temperature.
Kurihara, Akikazu; Yatabe, Toshio; Hiroi, Hiroshi; Tanabe, Hiromi
JNC TN9400 2003-060, 236 Pages, 2003/07
JNC-TN9400-2003-060.pdf:7.91MB
Large leak sodium-water reaction tests were carried out using SWAT-1 rig and SWAT-3 facility in Power Reactor and Nuclear Fuel Development Corporation (PNC) O-arai Engineering Center to obtain the data on the design of the prototype LMFBR Monju steam generator against a large leak accident.This report provides the results of SWAT-3 Runs 1 and 2.In Runs 1 and 2, the heat transfer tube bundle of the evaporator, fabricated by TOSHIBA/IHI, were used, and the pressure relief line was located at the top of evaporator.The water injection rates in the evaporator were 6.7kg/s and 14.2 (initial) - 9.7kg/s in Runs 1 and 2 respectively, which corresponded to 3.3 tubes and 7.1 (initial) - 4.8 tubes failure in actual size system according to iso-velocity modeling.Approximately two hundreds of measurement points were provided to collect data such as pressure,Temperature, strain,sodium level, void, thrust load, acceleration, displacement, flow rate, and so on in each run.Initial spike pressures were 1.13MPa and 2.62MPa nearest to injection point in Runs 1 and 2 respectively, and the maximum quasi-steady pressures in evaporator were 0.49MPa and 0.67MPa in Runs 1 and 2. No secondary tube failure was observed. The rupture disc of evaporator (RD601) burst at 1.1s in Run-1 and at 0.7s in Run-2 after water injected, and the pressure relief system was well-functioned though a few items for improvement were found.
Jitsu, Koji; Ono, Isao*; Kawada, Koji; Kurihara, Akikazu; Yatabe, Toshio
JNC TN9400 2003-062, 84 Pages, 2003/06
JNC-TN9400-2003-062.pdf:1.29MB
It is one of the important matters to select the design base leak (DBL) of the steam generator (SG) of a fast breeder reactor (FBR) in sodium-water reaction. The selection of the DBL has an influence on safety, economical efficiency, etc. of the plant.It is necessary to develop the computational model to estimate the sodium-water reaction phenomenon with high accuracy and rationality for selecting the DBL of large SGs. The blow down evaluation on overheating tube failure phenomenon is pointed out as part of the necessary improvements, since the behavior of overheating tube failure is largely affected by the steam conditions inside of the tube.This document shows the validation of blow down model for the LEAP code, which is developed as analysis code for failure propagation of the SG tubes, by test's data in Sodium-Water Reaction Test Rig No.1 (SWAT-1R). The following results have been obtained through the validation.(1) Within the mass flow rate ranging from 160 to 540 g/s in SWAT-1R, it has been confirmed that calculated internal pressure shows good agreement with the test's one. The pressure in tests becomes close to the calculation by Ogasawara model in critical flow models.(2) Mass flow rate is appropriately calculated of the test after about 10 seconds in the beginning of test. Calculated mass flow rate by Ogasawara model is closer to the test than that by Moody model.(3) On both Moody model and Ogasawara model in critical flow models, internal pressure in calculation shows the underestimation of HT-3 test. It will be necessary to investigate this reason.(4) Mass flow rate at the pipe near water heater tank shows the large overshoot in the beginning of test. Also, the small overshoot that supposed to be moved from upper stream appears at the nozzle about 2 seconds. Since these overshoots are estimated too large, these phenomena should require to be examined.(5) As calculated mass flow rate at the nozzle in both critical models have a tendency to be evaluated l
; ; ; Yatabe, Toshio
JNC TN9450 2001-004, 136 Pages, 2001/01
JNC-TN9450-2001-004.pdf:3.28MB
DNB (Departure from Nucleate Boiling) test were executed by a 1MWt Double-Wall-Tube Steam Generator. This data report describes the temperature fluctuation of the outer tube and sodium around DNB region. Furthermore, this report includes the temperature fluctuation of the inner surface of the inner tube obtained by removing noise of the original DNB signal and calculating heat flux of the tube, too. It also mentions the influence of the test parameter such as water flow rate on DNB period and DNB region length. All the DNB data described in this report were recorded by the data acquisition system of the small steam generator test facility. The contents are as followings : (1)1MW double-wall-tube steam generator, test method and test condition (2)The length of DNB region (3)DNB temperature fluctuation of the outer tube and sodium (experimental data) (4)The spectrum of DNB temperature fluctuation (5)DNB temperature fluctuation of the inner surface of the inner tube (calculation data)
; ; ; Yatabe, Toshio
JNC TN9400 2001-093, 88 Pages, 2001/01
JNC-TN9400-2001-093.pdf:2.35MB
The objectives of the 1MW test model of Double-Wall-Tube Steam Generator (DWT-SG) are to evaluate the thermal hydraulic performance and structural integrity such as heat transfer property, hydrodynamic instability property, DNB temperature fluctuation and capacity of leak detection for the elimination of the secondary sodium loop of FBR Plant. This report describes the heat transfer and hydrodynamic instability characteristic of DWT-SG on high water mass velocity (approximately 400
900 kg/m
sec) condition by tube plugging. Main results are followings : [Heat Transfer Characteristic] Evaluation of the tests clarified the sodium and water/steam side heat transfer coefficient and DNB quality correlations. Under normal output condition, the tube plugging induced no buckling of the tubes caused by temperature difference and no disproportionate sodium temperature distribution. (1) Sodium side heat transfer coefficient remains almost unchanged after tube plugging. Graber-Rieger's equation is better to be applied to this reigion than the other equations. (2) DNB quality of high mass velocity is also same as that of low mass velocity, Kon'kov's equation used for Phenix SG can predict experimental results in 20% errors. (3) As for pre-heat region, Dittus-Boelter's equation can predict experimental results in 20% errors in high mass velocity. But the heat transfer coefficient of this equation is larger than the experimental data in low mass velocity. (4) The experimental heat transfer coefficient of nucleate boiling region has a wide range, therefore it is very difficult to derive an experimental equation. Jens-Lottes' and Thom's equations are in the order of exprimental resaluts. (5) The range of the experimental heat transfer coefficient of film boiling region is a little wide, Mod. Tong's equation is better to be applied to this reigion than the other equations. (6) As for super-heat region, Bishop's equation can predict experimental results in 20% errors ...
; Yatabe, Toshio; ;
PNC TN9410 92-109, 71 Pages, 1992/04
The small test rig of double wall tube steam generator (DWTSG) was installed and experiments were started to evaluate the feasibility of DWTSG. Main objectives of this test rig are to obtain data on structural integrity of the double wall tube under steam generator circumstances such as DNB and instability flow and capability of leak detection. In order to study the basic thermal characteristics of the double wall tube steam generator, the tests for the thermal hydraulic characteristics and a preliminary DNB test were carried out from November, 1991. It is important to evaluate the data widely and timely, so preliminary results, analysis methods and results are described in this report. FOllowing results have been obtained so far. (1)The small test rig has a sufficient heat transfer capability. (2)Many kinds of thermal hydraulic data to evaluate the feasibility of DWTSG are obtained. (3)Test results shows the necessity of re-examination for the heat transfer calculation method at low load. (4)The structural integrity of the double wall tube caused by DNB must be maintained. We are going to carry out the experiments and evaluation.
; Yokochi, Yoji; *; ; Aoyama, Takafumi; ; Yatabe, Toshio
PNC TN9520 91-006, 861 Pages, 1991/07
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; ; Yatabe, Toshio
Technical committee meeting on "Sodium removal and disposal from LMFR's in normal operartion and in, ,
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Ida, Mizuho; Yoshida, Eiichi; Nakamura, Hiroo; Hirakawa, Yasushi; Yatabe, Toshio; Horiike, Hiroshi*; Kondo, Hiroo*; Yamaoka, Nobuo*
no journal, ,
no abstracts in English