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Mohamad, A. B.; Udagawa, Yutaka
Nuclear Technology, 210(2), p.245 - 260, 2024/02
Times Cited Count：1 Percentile：72.91(Nuclear Science & Technology)Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Yamaji, Akifumi*; Susuki, Naomichi*; Kaji, Yoshiyuki
IAEA-TECDOC-1921, p.199 - 209, 2020/07
The thermo-physical models and irradiation behavior of FeCrAl as defined by the benchmark organizer have been implemented to FEMAXI-7. Analyses were carried out firstly for the specified normal operation condition. Then, some sensitivity analyses were carried out with different assumptions and model parameters. Under the normal operating condition, the predicted FeCrAl cladded fuel performance was similar to that of Zry cladded fuel with notable, but not major difference regarding late gap closure. Under the simulated LOCA conditions, the burst pressure could be evaluated. The predicted cladding creep strain at burst was mainly attributed to creep strain with negligible plastic strain. Overall, FEMAXI-7 analyses have demonstrated excellent robustness and flexibility in modeling FeCrAl-UO
system under normal and LOCA conditions.
Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09
Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9
9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.
Suzuki, Motoe; Uetsuka, Hiroshi; Saito, Hiroaki*
Nuclear Engineering and Design, 229(1), p.1 - 14, 2004/04
Times Cited Count：19 Percentile：75.21(Nuclear Science & Technology)Mechanical load on cladding induced by fuel swelling in a high burn-up BWR type rod has been analyzed by a fuel performance code FEMAXI-6. The code has been developed for the analysis of LWR fuel rod behaviors in normal operation and transient conditions using FEM. During a power ramp for the high burn-up rod, instantaneous pellet swelling can significantly exceed the level that is predicted by a "steady-rate" swelling model, causing a large circumferential strain in cladding. This phenomenon has been simulated by a new swelling model to take into account the fission gas bubble growth, and as a result it has been found that the new model can give reasonable predictions on cladding diameter expansion in comparison with post-irradiation data. In addition, a pellet-clad bonding model which has been incorporated in the code to assume firm mechanical coupling between pellet outer surface and cladding inner surface has predicted the generation of bi-axial stress state in the cladding during ramp.
Suzuki, Motoe; Saito, Hiroaki*; Iwamura, Takamichi
Nuclear Engineering and Design, 227(1), p.19 - 27, 2004/01
Times Cited Count：7 Percentile：45.11(Nuclear Science & Technology)To assess the feasibility of the 31percentPu-MOX fuel rod design of reduced-moderation boiling water reactor in terms of thermal and mechanical behaviors, a single rod which is assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO fuel have been used complementally. The results are: FGR is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400K, while cladding diameter increase caused by pellet swelling is within 1percent strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling requires to be investigated in detail.
Suzuki, Motoe; Uetsuka, Hiroshi
IAEA-TECDOC-CD-1345 (CD-ROM), p.217 - 238, 2003/03
A fuel performance code FEMAXI-6 has been developed for the analysis of LWR fuel rod behaviors. The code uses FEM analysis, and has incorporated thermal and mechanical models of phenomena anticipated in high burn-up fuel rods. In the present study, PCMI induced by swelling in a high burn-up BWR type fuel rod has been analyzed. During a power ramp for the high burn-up fuel, instantaneous pellet swelling been simulated by a new swelling model which has been installed in the code to take into account the FP gas bubble growth, and the new model can give satisfactory predictions on cladding diametral expansion. In addition, a pellet-clad bonding model in the code, which assumes firm mechanical coupling between pellet outer surface and cladding inner surface, predicts an increased tensile stress in the axial direction of cladding during the power ramp, indicating the generation of bi-axial stress state in the cladding.
Uetsuka, Hiroshi
JAERI-Review 2002-027, 147 Pages, 2002/11
JAERI-Review-2002-027.pdf:9.54MB
The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the NSRR, the JMTR, the JRR-3 and the Reactor Fuel Examination Facility of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents.This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory.
Nakamura, Jinichi; Nakamura, Takehiko; Sasajima, Hideo; Suzuki, Motoe; Uetsuka, Hiroshi
HPR-359, Vol.2, p.34_1 - 34_16, 2002/09
In BWR, power oscillations can occur due to the void fraction fluctuation. To investigate the fuel behavior during power oscillation of BWRs, two types of irradiated fuel rods were tested under simulated power oscillation conditions in the Nuclear Safety Research Reactor(NSRR). One is high burnup BWR fuel (56GWd/t) test, with 4 power oscillation cycles, to clarify the behavior of high burnup fuel. The second one is high enriched fuel(20%,25GWd/t) test, with 7 power cycles, to perform the test under high power conditions. The fuel behavior data, such as cladding elongation, fuel stack elongation, cladding temperature, etc. were obtained in these tests. The DNB did not occur in these tests. The PCI was observed through cladding elongation and fuel stack elongation during the power oscillations, but the residual strain of cladding was very small. Fuel behavior under simulated power oscillations is discussed based on in-pile data and PIE data and is compared with FEMAXI-6 and FRAP-T6 calculation.
Suzuki, Motoe
Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.131 - 140, 2001/06
no abstracts in English
Lemehov, S.; Nakamura, Jinichi; Suzuki, Motoe
Nuclear Technology, 133(2), p.153 - 168, 2001/02
Times Cited Count：9 Percentile：56.08(Nuclear Science & Technology)no abstracts in English
Suzuki, Motoe
Nuclear Engineering and Design, 201(1), p.99 - 106, 2000/09
Times Cited Count：7 Percentile：46.88(Nuclear Science & Technology)no abstracts in English
Uetsuka, Hiroshi
JAERI-Review 2000-010, 113 Pages, 2000/07
JAERI-Review-2000-010.pdf:7.16MB
no abstracts in English
Lemehov, S.; Suzuki, Motoe
JAERI-Research 99-069, p.43 - 0, 2000/01
JAERI-Research-99-069.pdf:2.05MB
no abstracts in English
Saito, Hioraki*; Iriya, Yoshikazu*
JNC TJ8440 99-003, 156 Pages, 1999/03
no abstracts in English
; Saito, Hioraki*; *
Transactions of the 11th Int. Conf. on Structural Mechanics in Reactor Technology,Vol. C, p.1 - 6, 1991/08
no abstracts in English
Nakamura, Jinichi; *; Furuta, Teruo; *
JAERI-M 91-027, 36 Pages, 1991/03
no abstracts in English
Nakamura, Jinichi; Furuta, Teruo; *; *
HPR-339/13, 22 Pages, 1991/00
no abstracts in English
; Ki-S.Sim*
Res Mech., 25, p.101 - 128, 1988/00
no abstracts in English
; *
Nucl.Eng.Des., 101, p.267 - 279, 1987/00
Times Cited Count：25 Percentile：89.25(Nuclear Science & Technology)no abstracts in English
