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Sasa, Narimasa; Yamada, Susumu; Machida, Masahiko; Imamura, Toshiyuki*
Nonlinear Theory and Its Applications, IEICE (Internet), 7(3), p.354 - 361, 2016/07
A round off error accumulation in iterative use of the FFT is discussed. By using numerical simulations of partial differential equations, we numerically show that the round off error in iterative use of the FFT tend to be accumulated. To avoid a lack of precision, we give numerical simulations by using a quadruple precision floating point number, which ensure a sufficient precision against the round off errors by the FFT.
Nagai, Ryoji; Hajima, Ryoichi; Sawamura, Masaru; Nishimori, Nobuyuki; Kikuzawa, Nobuhiro; Minehara, Eisuke
Proceedings of 28th Linear Accelerator Meeting in Japan, p.381 - 383, 2003/08
no abstracts in English
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.).
*
Keisan Kogaku Koenkai Rombunshu, 1(1), p.113 - 116, 1996/05
no abstracts in English
*; Yokokawa, Mitsuo; Watanabe, Tadashi;
Keisan Kogaku Koenkai Rombunshu, 1(1), p.97 - 100, 1996/05
no abstracts in English
*; Yokokawa, Mitsuo; Watanabe, Tadashi;
JAERI-Data/Code 96-014, 22 Pages, 1996/03
JAERI-Data-Code-96-014.pdf:0.76MB
no abstracts in English
*
JAERI-Data/Code 96-007, 40 Pages, 1996/03
JAERI-Data-Code-96-007.pdf:1.36MB
no abstracts in English
Yokokawa, Mitsuo; Watanabe, Tadashi;
PERMEAN 95: Proc. of 1995 Int. Workshop on Computer Performance Measurement and Analysis, 0, p.54 - 59, 1995/00
no abstracts in English
E.A.Kenik*; Hojo, Kiichi
Journal of Nuclear Materials, 191-194, p.1331 - 1335, 1992/00
Times Cited Count：34 Percentile：92.8(Materials Science, Multidisciplinary)no abstracts in English
JAERI-M 5559, 46 Pages, 1974/02
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.
