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*; *; Kawasaki, Hirotsugu; Aoto, Kazumi
JNC TY9400 2000-010, 138 Pages, 2000/03
JNC-TY9400-2000-010.pdf:5.15MB
None
Kawasaki, Hirotsugu; *
JNC TN9400 2000-018, 37 Pages, 2000/03
JNC-TN9400-2000-018.pdf:1.34MB
The damage evaluation for the vessel model on the cyclic thermal transient loading in sodium were performed by the ultrasonic detection method. The wavelet analysis that was an analysis method of the waveform was applied to detect the micro damage before a sign of the crack initiation. The time-frequency analysis by the wavelet transform was performed to evaluate the ultrasonic parameter for the micro damage. As the results, the ultrasonic echo was analyzed by some mother wavelet, and Gabor wavelet was reasonable. The analysis of ultrasonic echo by Gabor wavelet showed drop of the sound velocity at higher frequency than the peak frequency because of attenuation in the high frequency component. The difference of the peak frequency △fp between B1 and B2 echoes increased with the amount of damage, and △ fp was available as a parameter for the micro damage detection. The correlation between the sound velocity and the micro hardness for the amount of damage was also found, and each method suggested to available alternately. ln this study, it was indicated that an ultrasonic wave characteristic value that can detect damaged state before crack initiation was obtained from the wavelet analysis.
; Kamide, Hideki;
PNC TN9410 98-083, 118 Pages, 1998/07
Large-scaled thermohydraulic tests are planned for new key technologies in the heat transport systems of a demonstration fast reactor. The test facility is consisted of components from a reactor vessel to a steam generator (SG). Basic design of the large-scaled thermohydraulic test facility is 1/3 scale of the demonstration fast reactor with two primary cooling loops and two into one secondary loop. The secondary piping length of the test facility is longer than the 1/3 scale of the demonstration fast reactor. The tests facility has the branch and junction of the secondary piping because of two primary loops and one SG. There is a possibility of flow and temperature unbalance if a buoyancy force were large and pressure loss were small. Therefore, dynamics analyses of the thermal transition tests had been done in which the secondary piping length. To examine the unbalance occurred or not, the natural circulation analysis had been performed providing different heat transfer area of the IHX or presser loss of the primary loop between A loop and B loop. It was shown from the analyses that the temperature response during the transition was delayed in the test model compared to the real reactor. Main cause of the delay was due to the real scaled SG. Other parameters, the length of piping etc., were not very influential to the response. The analysis such predicted that there wasn't large difference of global behaviors between the loops. Therefore, it was shown that there would be no problem, if the difference were made between the loops due to a manufacturing error.
kasahara, Naoto
PNC TN9520 93-009, 223 Pages, 1993/04
None
Muramatsu, Toshiharu
PNC TN9410 90-147, 115 Pages, 1990/10
In-vessel thermohydraulics analysis was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to evaluate effects of an inner barrel on a large fast breeder reactor. Then four thermohydraulics phenomena, a thermal stratification, a main loop temperature transient, a circumferential temperature distribution and a sodium surface velocity were discussed. Through the analysis using the multi-dimensional code AQUA and the discussion, the following have been effects of the inner barrel as obtained: [Thermal Stratification] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of an axial temperature distribution can be neglected from a structural design. [Main Loop Temperature Transient] An inner barrel is required. Because a cold shock with maximum temperature transient -2.0
C/s occurred at a outlet nozzle when an inner barrel was not equipped. [Circumferential Temperature Distribution] This phenomenon is not influenced very much on presence of an inner barrel. And appearance of the temperature distribution can be neglected from a structural design. But further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface velocity] An inner barrel is unnecessary. From the above results, it is concluded that an inner barrel is unnecessary if the cold shock is improved by a increase of effective mixing region on a design.
*; *; Uno, Tetsuro*
PNC TN9410 86-029, 68 Pages, 1986/02
A new test facility "Air-Cooling Thermal Transient Test Facility" (ATTF) was constructed at O-arai Engineering Center. This test facility is utilized, in the first place, for evaluating the strength of outlet tube-sheets of steam generators of FBR Plants. The objectives of the tube-sheet model tests are as follows. The first is to investigate and evaluate the strain concentration in the plastic region. The second is to confirm the adequacy of the design criteria for the prototype reactor MONJU. The third is to confirm the safety margin for failure incorporated in the design evaluation methods. ATTF can impose severe thermal loadings (only cold shock) on the test specimens. The facility produces compressed air (Max. 35kg/cm
G) by two large-sized compressors, and stores it in a storage tank (about 60m
). After a test specimen is heated up to the aimed temperature the compressed air passes through the test specimens quickly by opening the valve to apply cold shock and is released in the atmosphere. Each main loop pipe is 8 inches in diameter and the flow rate is max. 10kg/s in compressed air. The most severe down thermal transient condition is from 550C to 150C (for tube-sheet model) in about 4 min. The test section can be modified for various kinds of structures, which should be air-tight and have the maximum pressure of 8kg/cmG. The facility is operated automatically by two sequencer controllers. One of the main features of ATTF is the adoption of compressed air instead of sodium as coolant. By using compressed air, various kinds of sensors which can not be used in the sodium environment can be used in ATTF; particularly strain gages can be used effectively to obtain strain distribution for thermal transient condition, and the location as well as the mode of failure of test specimens can be recognized easily through the detection of crack initiation and the observation of crack growth. ATTF is expected to be a powerful ...
Tanaka, Masaaki; Hiyama, Tomoyuki; Murakami, Satoshi*; Doda, Norihiro; Ohshima, Hiroyuki
no journal, ,
In order to improve the accuracy of the numerical estimation method for the thermal transient phenomena in the sodium-cooled fast reactor has been conducted by using code coupling technology with the system analysis code for plant dynamics analysis, the multi-dimensional code for analysis of thermal-hydraulics in the plenum, and numerical estimation code of structural integrity for the local region, in viewpoint of enhancement of safety measures in sodium-cooled fast reactor. In this presentation, outline of simulation method development is introduced.
Chikazawa, Yoshitaka; Kubo, Shigenobu; Miyagawa, Takayuki*; Eto, Masao*
no journal, ,
Thermal transient on reactor vessel around sodium surface in a pool type reactor has been evaluated. Taking into account severe earthquake conditions, thickness of the reactor vessel is increased from the original design. Thermal transient evaluation including creep fatigue and ratchet at the sodium surface have been evaluated on that improved reactor vessel design against seismic load.