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Journal Articles

Study on loss-of-cooling and loss-of-coolant accidents in spent fuel pool, 1; Overview

Kaji, Yoshiyuki; Nemoto, Yoshiyuki; Nagatake, Taku; Yoshida, Hiroyuki; Tojo, Masayuki*; Goto, Daisuke*; Nishimura, Satoshi*; Suzuki, Hiroaki*; Yamato, Masaaki*; Watanabe, Satoshi*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

In this research program, cladding oxidation model in SFP accident condition, and numerical simulation method to evaluate capability of spray cooling system which was deployed for spent fuel cooling during SFP accident, have been developed. These were introduced into the severe accident codes such as MAAP and SAMPSON, and SFP accident analyses were conducted. Analyses using Computational Fluid Dynamics (CFD) code were conducted as well for the comparison with SA code analyses and investigation of detail in the SFP accident. In addition, three-dimensional criticality analysis method was developed as well, and safer loading pattern of spent fuels in pool was investigated.

Journal Articles

Behaviors of high-burnup LWR fuels with improved materials under design-basis accident conditions

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Journal Articles

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Takeda, Takeshi; Otsu, Iwao

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

 Times Cited Count:11 Percentile:80.31(Nuclear Science & Technology)

Journal Articles

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

Journal Articles

Development of the severe accident evaluation method on second coolant leakages from the PHTS in a loop-type sodium-cooled fast reactor

Yamada, Fumiaki; Imaizumi, Yuya; Nishimura, Masahiro; Fukano, Yoshitaka; Arikawa, Mitsuhiro*

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07

The loss-of-reactor-level (LORL) is one of the loss-of-heat-removal-system (LOHRS) of beyond-DBA (BDBA) severe accident. An evaluation method for the LORL which is caused by the coolant leakage in two positions of the primary heat transport system (PHTS) was developed for prototype JSFR which is loop-type sodium-cooled fast reactor. The secondary leakage in cold standby which occurred in different loop from that of the first leakage in rated power operation can lead LORL by excessive declining of the sodium level. Therefore, the sodium level behavior in RV was studied in a representative accident sequence by considering the sodium pumping up into RV, siphon-breaking to stop pumping out from RV and maintain the sodium level, and calculation programs for the transient sodium level in RV. The representative sequence with lowest sodium level was selected by considering combinations of possible leakage positions. As a result of the evaluation considering the countermeasures above, it was revealed that the LOHRS can be prevented by maintaining the sodium level for the operation of decay heat removal system, even in the leakages in two positions of PHTS which corresponds to BDBA.

Journal Articles

The Effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(11), p.1758 - 1765, 2016/11

 Times Cited Count:8 Percentile:61.89(Nuclear Science & Technology)

Journal Articles

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

 Times Cited Count:5 Percentile:41.49(Nuclear Science & Technology)

JAEA Reports

Safety demonstration test plan of the High Temperature Engineering Test Reactor (HTTR)

Tachibana, Yukio; Nakagawa, Shigeaki; Takeda, Takeshi; Saikusa, Akio; Furusawa, Takayuki; Takamatsu, Kuniyoshi; Nishihara, Tetsuo; Sawa, Kazuhiro; Iyoku, Tatsuo

JAERI-Tech 2002-059, 42 Pages, 2002/08

JAERI-Tech-2002-059.pdf:1.63MB

no abstracts in English

Journal Articles

Numerical investigation of heat transfer enhancement phenomenon during the reflood phase of PWR-LOCA

Onuki, Akira; Akimoto, Hajime

Journal of Nuclear Science and Technology, 36(11), p.1021 - 1029, 1999/11

 Times Cited Count:1 Percentile:13.4(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Application of simplified condensation model to PWR LBLOCA transient analysis with TRAC-PF1 code

; Murao, Yoshio

Journal of Nuclear Science and Technology, 33(4), p.290 - 297, 1996/04

 Times Cited Count:3 Percentile:33.18(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Applicability of REFLA/TRAC code to a small-break LOCA of PWR

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 32(3), p.245 - 256, 1995/03

 Times Cited Count:1 Percentile:17.86(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Assessment of predictive capability of REFLA/TRAC code for peak clad temperature during reflood in LBLOCA of PWR with small scale test, SCTF and CCTF data

; Onuki, Akira; Murao, Yoshio

Validation of Systems Transients Analysis Codes (FED-Vol. 223), 0, 8 Pages, 1995/00

no abstracts in English

JAEA Reports

Effect of fuel assembly configuration and fuel rod configuration on thermal-hydraulic behavior in core during reflood phase of PWR-LOCA

Onuki, Akira; ; Iguchi, Tadashi; Murao, Yoshio

JAERI-Research 94-012, 59 Pages, 1994/08

JAERI-Research-94-012.pdf:1.75MB

no abstracts in English

JAEA Reports

JAEA Reports

Assessment of one dimensional reflood model in REFLA/TRAC code

; Onuki, Akira; Murao, Yoshio

JAERI-M 93-240, 83 Pages, 1993/12

JAERI-M-93-240.pdf:1.94MB

no abstracts in English

JAEA Reports

Assessment of TRAC-BF1 1D reflood model with CCTF and SCTF data

; Onuki, Akira; Abe, Yutaka*; Murao, Yoshio

JAERI-M 93-045, 126 Pages, 1993/03

JAERI-M-93-045.pdf:2.56MB

no abstracts in English

JAEA Reports

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

; Onuki, Akira; Abe, Yutaka*; Murao, Yoshio

JAERI-M 93-028, 252 Pages, 1993/03

JAERI-M-93-028.pdf:5.98MB

no abstracts in English

Journal Articles

Thermal-hydraulic model for reflooding phenomena in a PWR-LOCA

Murao, Yoshio; Iguchi, Tadashi; Sugimoto, Jun; ; Iwamura, Takamichi; Okubo, Tsutomu; Onuki, Akira

Proc. of the 6th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics,Vol. 1, p.723 - 732, 1993/00

no abstracts in English

JAEA Reports

Study of two-phase flow under low velocity in PWR-LOCA

Onuki, Akira

JAERI-M 92-150, 134 Pages, 1992/10

JAERI-M-92-150.pdf:4.08MB

no abstracts in English

Journal Articles

Development of interfacial friction model for two-fluid model code against countercurrent gas-liquid flow limitation in PWR hot leg

Onuki, Akira; ; Murao, Yoshio

Journal of Nuclear Science and Technology, 29(3), p.223 - 232, 1992/03

no abstracts in English

68 (Records 1-20 displayed on this page)