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Journal Articles

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.

Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12

For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.

Journal Articles

Research on safety systems for future reactors using the ROSA/LSTF facility

Yonomoto, Taisuke; Otsu, Iwao; Nakamura, Hideo; Kondo, Masaya; Svetlov, S.*

Nippon Kikai Gakkai Dai-8-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.215 - 220, 2002/06

no abstracts in English

JAEA Reports

THYDE-NEU; Nuclear reactor system analysis code

Asahi, Yoshiro

JAERI-Data/Code 2002-002, 332 Pages, 2002/03

JAERI-Data-Code-2002-002.pdf:10.6MB

no abstracts in English

Journal Articles

A Spatial kinetics method ensuring neutronic balance with thermal-hydraulic feedback and its application to a main steam line break

Asahi, Yoshiro; Okumura, Keisuke; Ose, Yasuo*

Nuclear Science and Engineering, 139(1), p.78 - 95, 2001/09

 Times Cited Count:1 Percentile:86.33(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Program for steam table of heavy water for thermohydrodynamics analysis

Sato, Takeshi; Tamaki, Hitoshi

JAERI-Data/Code 2000-009, p.120 - 0, 2000/02

JAERI-Data-Code-2000-009.pdf:5.06MB

no abstracts in English

JAEA Reports

Development of input data for thermal-hydraulic computer code TRAC-BF1 for analyses of 1,100MW BWRs

*; Watanabe, Norio; Hirano, Masashi

JAERI-Data/Code 98-037, 193 Pages, 1998/11

JAERI-Data-Code-98-037.pdf:6.14MB

no abstracts in English

Journal Articles

Experimental analysis of coherent neutron flux fluctuations observed in a pressurized water reactor

Suzudo, Tomoaki; Tuerkcan, E.*; H.Verhoef*

Nuclear Science and Engineering, 129(2), p.203 - 208, 1998/06

 Percentile:100(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

B.Basoglu*; Yamamoto, Toshihiro; ; Nomura, Yasushi

JAERI-Data/Code 98-011, 89 Pages, 1998/03

JAERI-Data-Code-98-011.pdf:4.02MB

no abstracts in English

JAEA Reports

Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

Sato, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi; Iwamura, Takamichi; Murao, Yoshio

JAERI-Research 98-006, 77 Pages, 1998/02

JAERI-Research-98-006.pdf:3.27MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-034, 338 Pages, 1997/09

PNC-TN1410-97-034.pdf:6.65MB

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

Journal Articles

Research on core melt accident analysis of LWRs

Abe, Kiyoharu

Tokyo Daigaku Gakui Rombun, 0, 245 Pages, 1994/09

no abstracts in English

JAEA Reports

Safety analyses of a high conversion LWR with BWR type tight lattice core

Okubo, Tsutomu; Tomiai, Ichio*; Osugi, Toshitaka

JAERI-M 93-015, 72 Pages, 1993/02

JAERI-M-93-015.pdf:1.64MB

no abstracts in English

JAEA Reports

Analysis of Mihama-2 Steam Generator Tube Rupture(SGTR) event; Preliminary analysis

Hirano, Masashi; J.Sun*

JAERI-M 92-060, 61 Pages, 1992/04

JAERI-M-92-060.pdf:1.47MB

no abstracts in English

Journal Articles

Analyses of the Mihama-2 SGTR event and ROSA-IV experiment SB-SG-06 to simulate the event

Hirano, Masashi; Watanabe, Tadashi

Proc. of the 5th Int. Topical Meeting on Reactor Thermal Hydraulics: NURETH-5,Vol. 1, p.165 - 173, 1992/00

no abstracts in English

JAEA Reports

Thermal-hydraulic analysis of the Three Mile Island Unit 2 Reactor accident with THALES code

Hashimoto, Kazuichiro; Soda, Kunihisa

JAERI-M 91-193, 21 Pages, 1991/10

JAERI-M-91-193.pdf:0.63MB

no abstracts in English

JAEA Reports

None

*; *; *; *; *; *; *

PNC-TN1410 91-063, 239 Pages, 1991/08

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

no abstracts in English

JAEA Reports

Development of intellectual reactor design system; IRDS

; Nakakawa, Masayuki; Mori, Takamasa; ; *

JAERI-M 90-177, 96 Pages, 1990/10

JAERI-M-90-177.pdf:3.13MB

no abstracts in English

JAEA Reports

THYDE-W; RCS(reactor coolant system) analysis code

Asahi, Yoshiro; Matsumoto, Kiyoshi; Hirano, Masashi

JAERI-M 90-172, 305 Pages, 1990/10

JAERI-M-90-172.pdf:5.1MB

no abstracts in English

53 (Records 1-20 displayed on this page)