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Nagase, Fumihisa
Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.148 - 155, 2001/06
no abstracts in English
; ; Sakamoto, Naoki; *; Akasaka, Naoaki;
JNC TN9400 2000-095, 110 Pages, 2000/07
JNC-TN9400-2000-095.pdf:13.57MB
The effects of high fluence irradiation and swelling on the transient burst properties of austenitic steel fuel claddings; PNC316 and 15Cr-20Ni stcel, which were irradiated as the MONJU type fuel assemblies (MFA-1&MFA-2) in the FFTF reactor, were investigated. The temperature-transient-to-burst tests were conducted on a total of eight irradiation conditions. Fractographic examination and TEM observation were performed in order to evaluate the effect of high dose irradiation on the transient burst property and the relation between failure mechanism and microstructural change during rapid (ramp) heating. The results of the PIE showed that there was no significant effect of irradiation on the transient burst properties of these fuel claddings under the irradiation conditions examined. the results obtained in this study are as follows; (1)The rupture temperature of the irradiated PNC316 fuel cladding of MFA-1 was as same as that of our previous works for the fluence range up to 2.13
10
n/m
. There was no noticeable decrease in rupture temperature with increasing fluence in lower hoop stress region(
100MPa). (2)The rupture temperature of the irradiated 15Cr-20Ni fuel cladding of MFA-2 was almost as same as that of as-received cladding for the hoop stress range up to about 200MPa. The rupture temperature did not decrease significantly with fluence. (3)The rupture temperature of the irradiated PNC316 cladding tested at hoop stress 69MPa, which was the design hoop stress for MONJU fuel, was 1055.6
C. This suggested that the design cladding maximum temperature limit for MONJU (830C) was conservative. (4)There was no obvious relation between rupture temperature, swelling and microstructural change during transient heating under the irradiation conditions examined.
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.
Hamada, Hirotsugu
PNC TN9410 95-262, 35 Pages, 1995/09
Neighboring tubes in an FBR Steam Generator (SG) would suffer from overheating if a sodium-water reaction jet were formed due to water leakage in the SG. On the safety aspect of the SGs, it is important to confirm that the neighboring tubes would not fail under such an overheating condition. An analytical model using the structural integrity analysis code, FINAS, has been prepared to evaluate the overheating failure and here an explosive torch overheating test was analyzed to validate the FINAS model. These experiments and analysis indicate that the overheating failure is closely associated with heat transfer coefficients (HTCs) of outer and inner tube wall and that the FINAS model conservatively predicts the overheating failure within acceptable accuracy. For making progress in further tests like an explosive torch test and its code validation, it would be required that sodium-water reaction experiments should be performed to provide the data on the HTCs, high pressurized and superheated steam should be supplied in the explosive torch test, and that a multidimensional analytical model should be developed to closely predict the temperature distribution in the axial(z-) and circumferential(
-) directions on the tube wall.
Hamada, Hirotsugu; *; Matsuki, Takuo*
PNC TN9410 90-089, 150 Pages, 1990/05
A computer code SWACS (Sodium-Water Reaction Analysis Code System) was developed to analyze the pressure/fluid-flow phenomena during a large scale sodium-water reaction accident. Since the code was closely related with the prototype Monju SGs (steam generators) having a cover-gas space at their top, it didn't possess sufficient analytical functions for the specific phenomane of noncover-gas type SGs. In order to improve the accuracy of propagated pressure calculations and to make the code applicable to noncover-gas type SGs, some modifications related to new analytical models of pressure relief system, such as a dynamic response model of in-sodium RD (rupture disk) and a fluid-flow model in the piping connected to dump thak, were added to the calculation module of initial pressure spike and its propagation. The new version was named SWAC 57R. This report includes the results of analyses to verify the code functions and expermental validation to confirm code applicability to noncover-gas type SGs. The verification indicated the qualitative effects of calculation parameters (R
, R
, t-t, GAR, GVE) and the importance of dynamic fracture process of the RD as a moving variable orifice. The experimental validation using test date from PEPT series II showed that SWAC57R could accurately calculate propagated pressure (+20% at RD and +10% at IHX for the peak of pressure) by using property parameters (R = 0, R = 1, t-t = 0.004, GAR = 0.8, GVE = 0.5) Since the magnitude of the errors are comparable to that in the analysis of cover-gas type SGs conducted previously, it is concluded that the code can also be applied to noncover-gas type SGs.
JAERI-M 82-189, 33 Pages, 1982/11
no abstracts in English
Miura, Makoto; Hidaka, Yasuo; ; Omori, Takuro; Obata, Shinichi; Tanaka, Yasumasa; Shiina, Sadamu; Ogasawara, Koji
PNC TN841 79-12, 103 Pages, 1979/03
no abstracts in English
;
Genshiryoku Kogyo, 20(8), p.70 - 74, 1974/08
no abstracts in English
Narukawa, Takafumi; Amaya, Masaki
no journal, ,
no abstracts in English
Taniguchi, Yoshinori; Udagawa, Yutaka; Muramatsu, Yasuyuki; Hiruta, Kenta; Amaya, Masaki
no journal, ,
no abstracts in English
Ioka, Ikuo; Sakamoto, Kan*; Yamashita, Shinichiro
no journal, ,
For the purpose of technical base of new fuel cladding contributing to the safety improvement of LWR, the improved stainless steel (FeCrAl-ODS) fuel cladding superior in accident tolerance was developed. An integral LOCA examination of the FeCrAl-ODS fuel cladding was carried out at ORNL. It is confirmed that the tolerance for the LOCA burst of the FeCrAl-ODS fuel cladding improved substantially compared with Zircaloy.
