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Doda, Norihiro; Kato, Shinya; Uwaba, Tomoyuki; Tanaka, Masaaki; Nakamine, Yoshiaki*; Igawa, Kenichi*; Iida, Masaki*
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 14 Pages, 2025/08
Accurate evaluation of reactivity feedback due to core deformation during power increases in sodium-cooled fast reactors requires comprehensive modeling of the interactions among neutronics, thermal-hydraulics, and core mechanics. To accurately consider these interactions, JAEA has developed an evaluation method that combines multiple analysis codes that model these phenomena in detail. In this study, the evaluation method was applied to the core analysis of the FFTF LOFWOS Test #13, and the analysis results of net reactivity were compared with the test results. The sensitivity analysis results of the core structural design parameters showed that the core bowing behavior has a significant effect on the temporal variation of net reactivity.
Mizuta, Shunji; ;
JNC TN9400 99-082, 60 Pages, 1999/10
The density measurement of the internal creep specimens irradiated in FFTF/MOTA (Fast Flux Test Facility / Material open Test Assembly) was conducted MMF (Materia1 Monitoring Facility) and accurate separation of swelling strain from total strain leaded in the derivation of the irradiation creep coefficients. Irradiation creep coefficients for PNC 316, 15Cr-20Ni base S.S. and 14Cr-25Ni base S.S. were systematically expressed, while thermal creep coefficients K, under irradiation were separately expressed for above three steels. The results obtained are follows, (1)The effect of stress induced swelling was recognized in the temperature range from 405 to 605
C. The swelling in high stress specimens have a tendency to increasing swelling. (2)The irradiation creep coefficients derived from PNC316 and l5Cr-20Ni are similar to that of derived from 20%CW316S.S., CW316Ti and CW15-15Ti which were reported by other authors. (3)The irradiation creep coefficient derived from gas pressurized tube irradiation using FFTF/MOTA expressed appropriately irradiation creep strain from fuel pins using FFTF/MFA-2(15Cr-2ONi base S.S.).
E.A.Kenik*; Hojo, Kiichi
Journal of Nuclear Materials, 191-194, p.1331 - 1335, 1992/00
Times Cited Count：38 Percentile：93.45(Materials Science, Multidisciplinary)no abstracts in English
Ohgama, Kazuya; Takegoshi, Atsushi*; Hamase, Erina; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
no journal, ,
no abstracts in English
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
no journal, ,
Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.
Doda, Norihiro; Kato, Shinya; Yoshimura, Kazuo; Uwaba, Tomoyuki; Yokoyama, Kenji; Tanaka, Masaaki
no journal, ,
During a reactor power increase in ULOF and UTOP events in sodium-cooled fast reactors, core deformation due to thermal expansion of core elements is expected to cause a negative feedback effect to suppress this power increase. It is a complex phenomenon because of the interaction between reactor physics, thermal-hydraulics, and structural mechanics. To understand the phenomenon, JAEA has developed an evaluation method coupling the MARBLE code for reactor physics, the Super-COPD code for system-scale thermal-hydraulics, the ASFRE code for fuel-assembly-scale thermal-hydraulics, and the FINAS code for core structural deformation. This reports the sensitivity analysis results on core reactivity during a simulated ULOF test in FFTF as one of the validation tests for this evaluation method, and future development issues.
