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JAEA Reports

Decrease in coolability events analysis for the safety assessment of JRR-3 silicide core by THYDE-W code

Kaminaga, Masanori; Yamamoto, Kazuyoshi

JAERI-Tech 97-016, 120 Pages, 1997/03

JAERI-Tech-97-016.pdf:3.76MB

no abstracts in English

JAEA Reports

Reactivity initiated events analysis for the safety assessment of JRR-3 silicide core by EUREKA-2 code

Kaminaga, Masanori

JAERI-Tech 97-014, 125 Pages, 1997/03

JAERI-Tech-97-014.pdf:4.04MB

no abstracts in English

JAEA Reports

Reactivity initiated events analysis for the safety assessment of JRR-4 silicide LEU core

Kaminaga, Masanori; Yamamoto, Kazuyoshi; Watanabe, Shukichi; Nakano, Yoshihiro

JAERI-Tech 95-040, 79 Pages, 1995/07

JAERI-Tech-95-040.pdf:2.25MB

no abstracts in English

JAEA Reports

Safety analysis of JMTR-LEU cores, 1; Reactivity initiated accident analysis

Nagaoka, Yoshiharu; Komukai, Bunsaku; ; Saito, Minoru;

JAERI-M 92-095, 68 Pages, 1992/07

JAERI-M-92-095.pdf:1.52MB

no abstracts in English

JAEA Reports

Feedback control of primary circulation pump of PIUS-type reactor

*; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; *; Kukita, Yutaka

JAERI-M 91-076, 34 Pages, 1991/05

JAERI-M-91-076.pdf:1.02MB

no abstracts in English

Journal Articles

Safety characteristics of the High Temperature Engineering Test Reactor

Shindo, Masami; *; Kunitomi, Kazuhiko; ; Sawa, Kazuhiro

Nucl. Eng. Des., 132, p.39 - 45, 1991/00

 Times Cited Count:5 Percentile:53.89(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Feedback control of primary circulation pump of PIUS-type reactor during startup and steady state operation

*; Anoda, Yoshinari; Murata, Hideo; Yonomoto, Taisuke; Kukita, Yutaka; *

Thermal Hydraulics of Advanced Nuclear Reactors, p.85 - 89, 1990/11

no abstracts in English

Journal Articles

ROSA-IV large scale test facility(LSTF); Test program and first look of test results

Tasaka, Kanji; Koizumi, Yasuo

Nihon Genshiryoku Gakkai-Shi, 29(1), p.18 - 30, 1987/01

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

The Large Scal Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program is an integral test fcility to investigate thermal-hydraulic response of a pressurized water reactor (PWR) system during small break loss-of-coolant accidents (LOCAs) and operational transients.

Journal Articles

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

; ;

Nihon Genshiryoku Gakkai-Shi, 28(9), p.838 - 849, 1986/00

 Times Cited Count:0 Percentile:0.02(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

ROSA-IV Large Scale Test Facility(LSTF)System Description

*; Tasaka, Kanji; ; ; ; Koizumi, Yasuo; ; ; ; ; et al.

JAERI-M 84-237, 300 Pages, 1985/01

JAERI-M-84-237.pdf:7.57MB

no abstracts in English

JAEA Reports

Loss-of-Feedwater Transient Calculations for the ROSA-IV LSTF and the Reference PWR with RELAP5/MOD1(cycle 1)

C.P.Fineman*; ; Tasaka, Kanji

JAERI-M 83-088, 50 Pages, 1983/06

JAERI-M-83-088.pdf:1.5MB

no abstracts in English

JAEA Reports

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