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Matsui, Yoshinori; Takahashi, Hiroyuki; Yamamoto, Masaya; Nakata, Masahito; Yoshitake, Tsunemitsu; Abe, Kazuyuki; Yoshikawa, Katsunori; Iwamatsu, Shigemi; Ishikawa, Kazuyoshi; Kikuchi, Taiji; et al.
JAEA-Technology 2009-072, 144 Pages, 2010/03
JAEA-Technology-2009-072.pdf:45.01MB
"R&D Project on Irradiation Damage Management Technology for Structural Materials of Long-life Nuclear Plant" was carried out from FY2006 in a fund of a trust enterprise of the Ministry of Education, Culture, Sports, Science and Technology. The coupled irradiations or single irradiation by JOYO fast reactor and JRR-3 thermal reactor were performed for about two years. The irradiation specimens are very important materials to establish of "Evaluation of Irradiation Damage Indicator" in this research. For the acquisition of the examination specimens irradiated by the JOYO and JRR-3, we summarized about the overall plan, the work process and the results for the study to utilize these reactors and some facilities of hot laboratory (WASTEF, JMTR-HL, MMF and FMF) of the Oarai Research-and-Development Center and the Nuclear Science Research Institute in the Japan Atomic Energy Agency.
Ida, Mizuho; Nakamura, Hiroo; Chida, Teruo*; Miyashita, Makoto; Furuya, Kazuyuki*; Yoshida, Eiichi; Hirakawa, Yasushi; Miyake, Osamu; Hirabayashi, Masaru; Ara, Kuniaki; et al.
JAEA-Review 2008-008, 38 Pages, 2008/03
JAEA-Review-2008-008.pdf:9.37MB
Engineering Validation Design and Engineering Design Activity (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) is under going. IFMIF is an accelerator-based Deuterium-Lithium (D-Li) neutron source to produce intense high energy neutrons and a sufficient irradiation volume for testing candidate materials for fusion reactors. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid Li flow with a speed of 20 m/s. In target system, nuclear heating due to neutron causes thermal stress especially on a back-wall of the target assembly. In addition, radioactive species such as beryllium-7, tritium and activated corrosion products are generated. In this report, thermal stress analyses of the back-wall, mechanical tests on weld specimen made of the back-wall material, estimations of beryllium-7 behavior and worker dose at the IFMIF Li loop and consideration on major EVEDA tasks are summarized.
Okabe, Ayao; Onuki, Koji; Kikuchi, H.; M.Uchihashi; Nishibayashi, Yohei; Ikeda, Makinori; Miyake, Osamu
JNC TN2400 2003-005, 62 Pages, 2004/03
JNC-TN2400-2003-005.pdf:2.41MB
Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating validity of the mitigation system against secondary sodium leak of MONJU. The analytical results of floor temperature and hydrogen concentration were summarized in this report.In the sodium combustion analyses under the detailed design conditions, it was confirmed that the temperature rise of the floor liner was reduced In addition, as for the hydrogen concentration in sodium leak process which is formed by the reaction of sodium and moisture, it was confirmed that it is restricted under 4% of the hydrogen burn criterion. Concerning the hydrogen concentration due to the reaction with sodium and sodium hydroxide in the sodium pool after the storing, in the same way, it was confirmed that it is restricted under 4% of the criterion.
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.
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 ...
Ohno, Shuji; Matsuki, Takuo*; ; Miyake, Osamu
JNC TN9520 2000-001, 196 Pages, 2000/01
JNC-TN9520-2000-001.pdf:5.13MB
ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature.
; Miyake, Osamu;
PNC TN9410 98-037, 81 Pages, 1998/04
The sodium combustion computer code ASSCOPS has been developed for analyses of thermal consequences (i.e.pressure and temperature time histories) of sodium leak accidents in FBR plants. Version 2.0 of ASSCOPS, that is used in the study of this report, includes improvements and additional models over the previous versions. This report describes the validation of ASSCOPS (version 2.0) by using sodium pool combustion tests data obtained from FAUNA (F5, F6) at KfK, Germany, and SOLFA-1 (Run-D1) at PNC. The validation includes comparisons of calculation results of ASSCOPS (Version 2.0) with experimental data, and with calculation results of the previous version of ASSCOPS (Version 1.1). Furthermore, the effects of reaction products ratio (Na
O:NaO), initial humidity in the atomsphere, and radiation coefficient from the sodium pool to the gas were studied. The following results have been obtained from the study. (1)The calculation results agree well with the experimental data of the gas, sodium, and structure temperatures, and gas pressures. (2)The reaction products ratio (NaO:NaO) is one of the most important parameters for sodium combustion evaluation. It affects the pressure and temperature due to the difference of the reaction heat. Selection of proper value for this parameter results in the best estimate of the pressure, temperature and oxygen concentration. The ratio of NaO: NaO = 60: 40 is adequate for the purpose of conservative evaluation. (The analysis under the oxygen concentration below 10 % assumes NaO: NaO = 100: 0) (3)Initial humidity concentration in the air has been more little affect to the pressure and temperature than the reaction products ratio or the radiation coefficient of pool surface affect. (4)The radiation coefficient of pool surface was surveyed around the value obtained by conventional evaluation. The results shows that suppression of radiative heat transfer ...
; Ohno, Shuji; Miyake, Osamu; ; Seino, Hiroshi
PNC TN9520 97-001, 185 Pages, 1997/12
ASSCOPS(Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input, and output as the user's manual of ASSCOPS version 2.0. ASSCOPS is an integrated code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratoly in the U.S. The experimental studies conducted at PNC have been reflected in the ASSCOPS improvement. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (volume and structure surface area and thickness, etc.), and the atmospheric initial conditions, such as gas temperature, pressure, and gas composition. ASSCOPS calculates the time histories of atmospheric pressure and temperature changes along with those of the structural temperatures.
Nakagiri, Toshio; Miyake, Osamu; Ohno, Shuji
PNC TN9410 97-102, 166 Pages, 1997/11
The calculation of Sodium Fire Test - I (Run - E6) was performed using the ASSCOPS (Analysis of Simultaneous Sodium Combustions in Pool and Spray) code version 2.0 in order to determine the parameters used in the code for the calculations of sodium combustion behavior of small or medium scale sodium leak, and to validate the applicability of the code. The parameters used in the code were determined and the validation of the code was confirmed because caluculated temperatures, calculated oxygen concentration and other calculated values almost agreed with the test results.
; ; 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 ...
Miyake, Osamu; ;
Nuclear Technology, 0 Pages, 1995/01
A general framework to represent and characterize dominant in-vessel and ex-vessel sequences ofan LMFBR severe accident is presented with emphasis on the role of physical barriers and relevant mitigation features against major loading mechanisms.An assessment of ULOF-initiated accident spectra for a model plant identified the risk-dominant seaquences and their