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Journal Articles

Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04

A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.

Journal Articles

Present status of thermal design of nuclear reactors using large-scale simulations

Takase, Kazuyuki

Nihon Kikai Gakkai Ryutai Kogaku Bumon Nyuzu Reta Nagare (Internet), 6 Pages, 2005/04

no abstracts in English

Journal Articles

Large-scale simulations on bubbly flow dynamics in a fuel channel with the earth simulator

Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*

Hai Pafomansu Komputingu To Keisan Kagaku Shimpojium (HPCS 2005) Rombunshu, P. 16, 2005/01

no abstracts in English

Journal Articles

Recent activities on subchannel analysis at JAERI

Okubo, Tsutomu; Araya, Fumimasa; Iwamura, Takamichi; Kusunoki, Tsuyoshi

Fourth Int. Seminar on Subchannel Analysis (ISSCA-4), p.267 - 286, 1997/00

no abstracts in English

JAEA Reports

None

*; Kinjo, Hidehito*; *; Ito, Kunihiro*; *

PNC TJ1214 92-007, 105 Pages, 1992/07

PNC-TJ1214-92-007.pdf:2.82MB

None

JAEA Reports

None

*; *; Fukumura, Nobuo*; *; *; *; *

PNC TN1410 91-063, 239 Pages, 1991/08

PNC-TN1410-91-063.pdf:10.66MB

no abstracts in English

Journal Articles

A Study on Hot Spot Factor and on the Degree of Uncertainty in Thermal Design of Reactor Core

Nihon Genshiryoku Gakkai-Shi, 12(4), p.179 - 186, 1970/00

no abstracts in English

Oral presentation

Core design for the next generation sodium-cooled fast reactor, 2; Reference core design

Tsuboi, Toru*; Moriwaki, Hiroyuki*; Ogura, Masashi*; Hibi, Koki*; Maeda, Seiichiro; Ohgama, Kazuya; Chikazawa, Yoshitaka; Oki, Shigeo

no journal, , 

no abstracts in English

Oral presentation

Design of a low sodium void reactivity core concept of tank-type sodium-cooled fast reactor

Hasegawa, Takashi*; Kan, Taro*; Moriwaki, Hiroyuki*; Tokizaki, Minako*; Yamano, Hidemasa; Takano, Kazuya

no journal, , 

no abstracts in English

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