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Ito, Kazuhiro; Tanabe, Hiromi; Kaneko, Yoshihisa; Kagota, Eiichi; Takahashi, Yasuo
Proceedings of 2013 International Congress on Advances in Nuclear Power Plants (ICAPP 2013) (USB Flash Drive), 10 Pages, 2013/04
The Monju is equipped with two types of hydrogen-meters to detect water leakage in steam generators. Since they are so highly-sensitive as to detect minor water leak from a steam generator tube, they sometimes detect hydrogen concentration increases at plant operational condition changes such as start-up without any water leak. No water leak was experienced during one year operation of the Startup Test up to 40% in 1995, although hydrogen concentration sometimes increased at plant operational condition changes. The H behavior of Monju IHTS during the previous Startup Test was examined and the following knowledge was obtained: The in-sodium H behaves in parallel with the IHTS sodium temperature. In-cover-gas H behavior is more complicated and sensitive to plant operational condition changes such as plant load changes than the in-sodium one. Both types of H-meters underwent a certain degree of zero level drift during one year operation.
Yanagisawa, Tsutomu; Tanabe, Hiromi
Nihon Genshiryoku Gakkai-Shi ATOMO
, 49(7), p.499 - 504, 2007/07
The first story of this serial handles the early days of FBRs. The progress of nuclear physics in the 1930's led to the development of the FBR. Fast neutrons as for fission generator, plutonium for nuclear fuels and liquid metal as for the coolant were selected to obtain a higher breeding ratio. Based on the experience of the Clementine Reactor, the EBR-I went critical in August 1951. Moreover, they succeeded in electricity generation in December 1951 and later confirmed breeding. The results of EBR-I project are thought to demonstrate the controllability of the fast neutron, the utility of liquid metals, and the reality of breeding.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi
JNC TN2400 2003-003, 225 Pages, 2004/02
JNC-TN2400-2003-003.pdf:40.45MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies.
Hamada, Hirotsugu; Tanabe, Hiromi
Journal of Nuclear Science and Technology, 41(6), 665 Pages, 2004/00
Times Cited Count：1 Percentile：10.03(Nuclear Science & Technology)Sodium-water reaction tests simulating intermediate water leaks into sodium are analyzed by the overheating rupture model of a heat-transfer tube exposed to the corrosive and high-temperature reaction jet. The comparison of the model with test data leads to the following conclusions: The failure behaviors of gas-pressurized tubes are classified into two types of the creep failure and the ductile failure accompanied by creep, depending on the test conditions. Thin-wall tubes fail within tens of seconds due to the ductility while thick-wall tubes fail in about 1 minute due to the creep; in the latter case, the thinning due to wastage proved to be dominant in the overheating rupture phenomena. In the creep failure and the ductile failure accompanied by creep, the times to failure in the analysis are respectively estimated 20-50 % and 35-50 % shorter than those in the experiments; namely, the analysis has a certain degree of conservatism. In the creep failure, if the time coefficient
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.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; ; Miyakawa, Akira; Okabe, Ayao;
JNC TN9400 2001-130, 235 Pages, 2002/03
JNC-TN9400-2001-130.pdf:7.05MB
The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: (1)To evaluate the structural integrity of tube material, the strength standard for 2.25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200
C) creep data. This standard has been validated with the tube rupture simulation test data. (2)The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. (3)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. (4)The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. (5)The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system.
Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Okabe, Ayao; Miyakawa, Akira
JNC TN9400 2001-099, 76 Pages, 2001/11
JNC-TN9400-2001-099.pdf:2.13MB
The JNC technical report "The Development and Application of overheating Failure Model of FBR Steam Generator Tubes" summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. (1)0n the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. (2)For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. (3)Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure.
; Takai, Toshihide; ; ; Miyake, Osamu; Tanabe, Hiromi
JNC TN9400 2000-090, 413 Pages, 2000/08
JNC-TN9400-2000-090.pdf:16.61MB
As a part of the work for investigating the sodium leak accident which occurred in the Monju reactor (hereinafter referred to as Monju), sodium fire test-II was carried out using, the SOLFA-1 (Sodium Leak, Fire and Aerosol) facility at OEC/PNC. ln the test, the piping, ventilation duct, grating and floor liner were all full-sized and arranged in a rectangular concrete cell in the same manner as in Monju. The main objectives of the test were to confirm the leak and burning behavior of sodium from the damaged thermometer, and the effects of the sodium fire on the integrity of the surrounding structure. The main conclusions obtajned from the test are shown below: (1)Burning Behavior of Leaked Sodium : lmages taken with a cameras in the test reveal that in the early stages of the sodium leak, the sodium dropped down out of the flexible tube in drips. (2)Damage to the ventilation Duct and Grating: The temperature of the ventilation duct's inner surface fluctuated between approximately 600C and 700C. The temperature of the grating began rising at the outset of the test, then fluctuated betvveen roughly 600C and 900C. The maximum temperature was about 1000C. After the test, damage to the ventilation duct and the grating was found. Damage to the duct was greater than that at Monju. (3)Effects on the Floor Liner : The temperature of the floor liner under the leak point exceed l,000C at 3 hours and 20 minutes into the test. A post test inspection of the liner revealed five holes in an area about 1m 
1m square under the leak point. There was also a decrease of the liner thickness on the north and west side of the leak point. (4)Effects on Concrete: The post test inspection revealed no surface damage on either the concrete side walls or the ceiling. However, the floor concrete was eroded to a maximum depth 8 cm due to a sodium-concrete reaction. The compressive strength of the ...
Kawada, Koji; ; Ohno, Shuji; ; Miyake, Osamu; Tanabe, Hiromi
JNC TN9400 2000-089, 258 Pages, 2000/08
JNC-TN9400-2000-089.pdf:12.26MB
As a part of the work for investigating the sodium leak accident which occurred in the Monju reactor (hereinafter referred to as Monju) on December 8, 1995, threetests, (1)a sodium leaktest, (2)a sodium leak and fire test-I, and(3)a sodium leak and fire test-II, were carried out at OEC/PNC, The main objectives of these tests were to confirm the leak and burning behavior of sodium from the damaged thermometer, and the effects of the sodium fire on the integrity of the surrounding structure. This report describes the results of the sodium fire test-I carried out as a preliminary test. The test was performed usjng the SOLFA-2 (Sodium Leak, Fire and Aerosol) facility on April 8, 1996. In this test, sodium heated to 480C was leaked for approximately l.5 hours from a leak simulating apparatus and caused to drop onto a ventilation duct and a grating with the same dimensions and layout as those in Monju. The main conclusions obtained from the test are shown below: (1)Observation from video cameras in the test revealed that jn the early stages of the sodium leak, sodium dripped out of the flexible tube of the thermometer. This dripping and burning expanded in range as the sodium splashed on the duct. (2)No damage to the duct itself was detected. However, the aluminum louver frame of the ventilation duct's lower inlet was damaged. lts machine screws came off, leaving half of the grill (on the grating side) detached. (3)NO large hole, like the one seen at Monju, was found when the grating was removed from the testing system for inspection, although the area centered on the point were the sodium dripped was damaged in a way indicating the first stages of grating failure. The 5mm square lattice was corroded through in some parts, and numerous blades (originally 3.2 mm thick) had become sharpened like the blade of a knife. (4)The burning pan underside thermocouple near the leak point measured 700C in within approximately 10 minutes, and for the next ...
Hamada, Hirotsugu; *; *; *; Hiroi, Hiroshi*
PNC TN9410 98-029, 122 Pages, 1998/05
The following items have been studies to evaluate overheating failure of FBR generator heat transfer tubes: (1)To establish a structural integrity analysis method. The strength standard values for 2.25Cr-1Mo steel was established taking account of time dependent effect to overheating failure mechanism based on high temperature (700 - 1200C) creep data and was validated by tube rupture simulation test data. (2)To improve and validate blow down analytical method. The analytical result by use of BLOOPH, the FBR blow down code, was compared with that by use of RELAP-5, the general purpose thermo-hydraulic code, and a good agreement was obtained. (3)To quantitatively validate the entire overheating analysis model by sodium water reaction data Sodium-water reaction tests of SWAT-3 and LLTR were analyzed using above mentioned analytical method. The ductile fracture occurred earlier than the creep fracture in the analysis and the comparison of tube failure times with the experiments showed sufficient conservativeness. Based on the above studies, the analytical method was applied to PFR superheater leak event and the Monju steam generator accidental analysis. The followings were quantatitively shown through the analysis: (1)The most important cause that multi-tube failure occurred in the 1987 PFR superheater-2 leak is that the superheater did not equip a fast steam dump system at the time of the leak event. (2)Overheating failure will not occur under any operational conditions of Monju in both steady state and transient phases such as water/steam blow-down. (3)Although safety margin becomes small when the water/steam flow rate becomes small during the blow-down, the modification of the plant such as hastening blow-down by equipping more relief valves will drastically improve the safety margin.
; ; Tanabe, Hiromi; Ohno, Shuji; Miyake, Osamu;
PNC TN9410 97-030, 93 Pages, 1997/04
A sodium fire analysis code, ASSCOPS(Analysis of Simultaneous Sodium Combustions in Pool and Spray) was developed coupling the computer codes of SPRAY-IIIM and SOFIRE-MIl to assess temperature-pressure transients resulting from sodium spray and pool combustions, simultaneously. The validation of ASSCOPS was conducted using the experimental results obtained from sodium spray fire experiments using 21 m
vessel and the accuracy of calculated results was discussed. The following results were obtained: (1)Study under inert gas atmosphere. The comparison between analysis and experiment with regard to the pressure and the temperature showed a good agreement. (2)Study under air atmosphere. The comparison between analysis and experiment with regard to the pressure and the temperature also showed a good agreement. (3)Effects of parameter used in evaluating the design of Monju. The peak pressure and temperature obtained by the analysis overestimates the experimental results. From these results, it was concluded that the development and validation of ASSCOPS indicate a improvement on the burning and the heat transfer models in SPRAY-IIIM.
Shimoyama, Kazuhito; Usami, Masayuki; Miyake, Osamu; ; ; Tanabe, Hiromi
PNC TN9450 97-007, 81 Pages, 1997/03
None
Uchiyama, Naoki; Takai, Toshihide; Nishimura, Masahiro; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi
PNC TN9410 97-051, 383 Pages, 1997/03
As a part of the work for investigating the sodium leak accident which occurred in Monju, sodium fire test-II was carried out using SOLFA-1 (Sodium Leak, Fire and Aerosol) facility at OEC/PNC. In the test, the piping, ventilation duct, grating and floor liner were an full-sized and arranged in a rectangular concrete cell in the same manner as Monju. Main objectives of the test are to confirm leak and burming behavior of sodium from the damaged thermometer, and effects of the sodium fire on integrity of the surrounding structure, etc. The main conclusions obtained from the test are shown as below. (1)Burning Behavior of Leaked Sodium : Images taken with cameras in the test reveal that in the early stages of the sodium leak, the sodium dropped down out of the flexible tube in drips. This dripping and burning were expanded in range as the sodium splashed on the duct, the grating and a support of thermocouples for the measurement of gas temperature. (2)Damage of Ventilation Duct and Grating : The temperature of the ventilation duct's inner surface fluctuated between approximately 600C and 700C. The temperature of the grating began rising at the outset of the test, then fluctuated between roughly 600C and 900C. The maximum temperature was about 1000C. After the test, damage of the ventilation duct and the grating was found. Damage of the duct was greater than that of Monju. (3)Effects on Floor Liner : The temperature of the floor liner under the leak point exceeded 1,000C at 3 hours and 20 minutes of the test; Post test inspection of the liner revealed five holes in the region of about 1m 1m under the leak point. There was also a decrease of a liner thickness on the north side and west side of the leak point. (4)Effects on Concrete : No surface damage of the concrete side walls and ceiling was found by the post test inspection. The floor concrete was eroded by a depth of 8 cm at maximum due to ...
Kawada, Koji; Ohno, Shuji; Miyake, Osamu; ; ; Tanabe, Hiromi
PNC TN9410 97-036, 243 Pages, 1997/01
As a part of the work for investigating the sodium leak accident which occurred in Monju on December 8, 1995, three tests, (1)sodium leak test, (2)sodium leak and fire test-I, and (3)sodium leak and fire test-II, were carried out at OEC/PNC. Main objectives of these tests are to confirm leak and burning behavior of sodium from the damaged thermometer, and effects of the sodium fire on integrity of the surrounding structure, etc. This report describes the result of the sodium fire test-I carried out as a preliminary test. The test was performed using SOLFA-2 (Sodium Leak, Fire and Aerosol) facility on April 8, 1996. In this test, sodium heated to 480C was leaked for approximately 1.5 hours from a leak simulated apparatus and caused to drop onto a ventilation duct and a grating with the same dimensions and layout as those in Monju. The main conclusions obtained from the test are shown as below. (1)Observation from video cameras in the test revealed that in early stages of sodium leak, sodium dropped down out of the flexible tube of thermometer in drips. This dripping and burning were expanded in range as sodium splashed on the duct. (2)No damage to the duct itself was detected. However, the aluminum louver frame of the ventilation duct's lower inlet was damaged: Its machine screws had come off, leaving half of the grill (on the grating side) detached. (3)No large hole, like one seen at Monju, were found when the grating was removed from the testing system for inspection, although the area centered on the point that the sodium attacked was damaged in a way indicating the first stages of grating failure: The 5-mm- square lattice was corroded through in some parts, and many blades (originally 3.2 mm thick) had become like the blade of a sharp knife. (4)The burning pan underside thermocouple near the leak point measured 700C in roughly 10 minutes, and for the next hour remained stable between 740C and 770C. There was a ...
Shimoyama, Kazuhito; Nishimura, Masahiro; Usami, Masayuki; Miyahara, Shinya; Miyake, Osamu; Tanabe, Hiromi
PNC TN9410 97-085, 163 Pages, 1996/11
Sodium fire experiments were carried out two times using the Sodium Fire Test Rig (SOFT-1) in the Power Reactor and Nuclear Fuel Development Corp (PNC) as a part of works to research the cause of the accident in secondary main cooling system of Monju. The purposes of these experiments are to confirm the leak rate and leakage form of sodium from damaged thermometer, to confirm the damage to the piping insulating structure around the thermometer and to the flexible tube, and to compare the temperature history of the signal from the thermometer between the experiments and Monju. In the experiments 56(
2)g/sec was obtained as the leak rate under the condition of ensuring the leakage pass in the simulated thermometer. This leak rate was corrected to 53g/sec to take account of manufacturing error of the theemometer between the experiment and Monju. In calculation of this leak rate, it is assumed that the annulus size of thermometer well tip is a nominal distance and pressure value to the leakage sodium is 1.65kg/cm
G, which was the maximum one during the leakage of Monju. Concerning the leakage form, connection condition between the thermometer and flexiblc tube affected the dropping style of the leaking sodium especially in its initial behavior. For the connection condition of the thermometer and flexible tube at the beginning of the experiments, the first experiment was started removing the connection to simulate the post accident observation results of Monju, while the second one was started in connected condition. In the second experiment, the connection condition became to be equal with the initial state of the first experiment 17 seconds after the beginning of thc leak ; the cap nut which fixed the flexible tube to the elbow connector came off. Until the connection came off, the typical leakage form was the dispersion from the elbow connector as a droplet and the flow penetrating the covering of the flexible tube as a streamline, while after the ...
Hamada, Hirotsugu; Tanabe, Hiromi
PNC TN9410 96-027, 41 Pages, 1995/12
If a sodium-water reaction jet was formed due to water leakage in an FBR steam generator(SG), neighboring tubes would suffer from overheating. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such a severe overheating condition. So far, an analytical model using the structural integrity analysis code, FINAS, has been prepared and validated by the explosive torch overheating test data. This report presents the results on the overheating failure analysis of the under-sodium leak in the PFR superheater(SH), 1987. In the SH with slow steam dump system in 1987, neighboring overheated tubes are failed about 3 seconds after the SH isolation, which is shown both by the leak in the PFR and its analysis. For the SH in which a fast steam dump system was installed after the leak of 1987, the analysis shows no tube failure due to the fast steam depression and cooling effect inside. These results indicate that the FINAS model adequately predicts the overheating failure and the specific SH design and operation possibly result in further growth of the leak. It is concluded that steam blow effect is extremely important and the analysis model presented here is useful for the overheating failure evaluation of the SGs.
