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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 - 1200
C) 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.
; 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 ...
*; Ozawa, Kenji; Yoshikawa, Shinji; *
PNC TY1602 97-001, 36 Pages, 1997/04
no abstracts in English
Takeda, Toshikazu*; *; Kitada, Takanori*; *
PNC TJ9605 97-001, 100 Pages, 1997/03
This report is composed of the following two parts and appendix. (I)Improvement of the Method for Evaluating Reactivity Based on Monte Carlo Perturbation Theory (II)Improvement of Nodal Transport Method for 3-D Hexagonal Geometry (Appendix) Effective Cross Section of 
U Samples for Analyzing Doppler Reactivity in Fast Reactors Part I. Improvement of the Method for Evaluating Reactivity Bascd on Monte Carlo Perturbation Theory. Theoretical formulation in Monte Carlo perturbation method had been checked, and then introduced into a calculation code. The increase of CPU time is about 10 to 20 % compared to that if normal Monte Carlo code, in the cases of same number of history. This Monte Carlo perturbation method found to be effective, because results are almost reasonable and deviations of the results are especially small, by using the Monte Carlo perturbation code. However, there are somc cases that the results of the change of eigenvalues becomes positive or negative by changing the estimator, and there is no reasonable difference in the results between the conventional method, which does not consider the change of neutron source distribution caused by a perturbation, and the new method, which consider that change. Thus it is still necessary to check the Monte Carlo pcrturbation code. Part II. Improvement of Nodal Transport Method for 3-D Hexagonal Geometry It is certain that we can accurately evaluate hexagonal geometry FBR core by nodal transport calculation code for hexagonal-Z geometry named 'NSHEX'. However it is also found that in very heterogeneous core the results is not good enough. Because the treatment of the transverse leakage to the radial direction, which is use for evaluating intra-nodal flux distribution, is not so accurate. For the treatment of the leakage distribution, it is necessary to estimate the nodal vertex flux. In conventional method, the vertex flux estimated by the surrounding node surfacc flux around that vertex. On the contrary,
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 ...
Aoto, Kazumi; ;
PNC TN9410 97-037, 51 Pages, 1996/11
The basic material properties of a rolled steel for welded structure (present standard name is SM400B, old standard name SM41B) which is used as the liner plate in SHTS cells of "Monju plant". Based on the material testing data for evaluation of structural integity of the liner during sodium leakage are tentatively proposed. Main basic material properties are shown as follows. (1)The 0.2% offset yield stress (lower yield point). (2)The ultimate tensile strength. (3)The modulus of the longitudinal elasticity. (4)Static stress-strain relation. (Physical property in Ludwik equation). (5)The creep strain. (6)The linear thermal expansion coefficient. (7)The density. (8)A specific heat. (9)The thermal conductivity.
Aoto, Kazumi; ; Hirakawa, Yasushi
PNC TN9410 97-055, 128 Pages, 1996/07
Several material analyses on damage of the floor liner made of a mild steel which was in the test cell of the second sodium leak and combustion experiment (Test-2) performed in OEC/PNC on June 7 in 1996 were carried out to clarify the following issues. (1)Difference of the corrosion mechanism of Test-2 liner to that of the first sodium leak and combustion experiment(Test-1) liner. (2)The vital factor which can desides corrosion mechanism and damage location. The following analyses were accomplished. (a)Microstructure observation (b)EPMA for cross-section of vicinity of corroded area (c)X-ray diffraction(XRD) for the interface between corrosion product-liner(mild steel) The differences between the corrosion mechanism of Test-1 liner which is seemed to be the same that of "MONJU" liner and that of Test-2 liner is discussed based on the results of these material analyses. As the result, the Na-Fe double oxidization with mechanical/chemical removal of reaction product can be occurred on the Test-1 and "MONJU" liner. On the other hand, a hot-corrosion, taht is the molten salt type corrosion is subject to be thinning of the Test-2 liner. All failures of Test-2 liner surround at the halfway up a convex. Considering the above corrosion mechanism, that fact leads that significant damage is occurred at the molten salt level.
Aoto, Kazumi; ; ; ; ; Hirakawa, Yasushi
PNC TN9410 97-076, 29 Pages, 1996/06
This report describes the check of the thermocouple well at the inlet of the intermediate heat exchanger (IHX) of C-loop of the secondary heat tansfer system of the prototype fast breeder reactor Monju, regarding the sodium leak accident of the thermocouple well at the outlet of the IHX of the same loop of the secondary heat transfer system of the same plant Monju. Various tests and inspections were performed to check damages at the part with rapid diameter change of the thermocouple well where stress concentration may occur, and to get the information about the integrity of the welded part between the thermocouple well and the attachment. The thermocouple well, the rapid diameter change part, larger and smaller part, respectively, of the thermocouple well, and welded part between the thermocouple well and the attachments were examined as written below. (1)Accurate measurement of the dimension. (2)Vibration test by tapping the thermocouple well. (3)Non destructive testing at some points. (4)Chemical composition analysis. (5)Microscopic observation of metalogical structure. (6)Detailed observation around the rapid diameter change part. (7)Hardness test. (8)Research on corrosion at the clearance. (9)Structure strength test of the thermocouple well. (10)Bending test of the thermocouple's sheath at high temperature.
Ohtaki, Akira;
PNC TN9410 96-142, 102 Pages, 1996/04
In order to evaluate the pressure history at the leaked position and the amount of sodium leakage regarding the sodium leak accident of MONJU, the plant system dynamics were calculated by Super-COPD. It was estimated from the calculated results that the sodium leakage was halted around 23:28, which was 3 hours and 41 minutes after the initiation. The pressure loss coefficient of the leaked position was evaluated to be 2.16 in two kinds of the water experiments of PNC and of Toshiba-IHI. Using cofficient to the calculated pressure history, the minimum and the maximum leakage rates were evaluated to be 35.5 and 51.9g/sec respectively, and the average rate was 48.9g/sec. Therefore, the total amount of sodium leakage was estimated to be 650
38kg.
