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Journal Articles

New reactor cavity cooling system (RCCS) with passive safety features; A Comparative methodology between a real RCCS and a scaled-down heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*

Annals of Nuclear Energy, 96, p.137 - 147, 2016/10

 Times Cited Count:4 Percentile:50.65(Nuclear Science & Technology)

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.

Journal Articles

New reactor cavity cooling system with a novel shape and passive safety features

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Morita, Koji*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1250 - 1257, 2016/04

After Fukushima Daiichi nuclear disaster by TEPCO, a cooling system to prevent core damage became more important from the perspective of defense in depth. Therefore, a new, highly efficient RCCS with passive safety features without a requirement for electricity and mechanical drive is proposed. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. The RCCS can always stably and passively remove a part of the released heat at the rated operation and the decay heat after reactor shutdown. Specifically, emergency power generators are not necessary and the decay heat can be passively removed for a long time, even forever if the heat removal capacity of the RCCS is sufficient. We can also define the experimental conditions on radiation and natural convection for the scale-down heat removal test facility.

Journal Articles

New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

Takamatsu, Kuniyoshi; Hu, R.*

Annals of Nuclear Energy, 77, p.165 - 171, 2015/03

 Times Cited Count:11 Percentile:20.79(Nuclear Science & Technology)

A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed. The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 ($$^{circ}$$C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal.

Journal Articles

Containment pressure suppression system with functions of water injection and noncondensable gas confinement

Yonomoto, Taisuke; Okubo, Tsutomu; Iwamura, Takamichi; Ishida, Toshihisa

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 8 Pages, 2003/04

An innovative concept of the pressure suppression system having functions of water injection and non-condensable gas confinement is developed for the next generation light water reactors (LWRs). The use of the system is advantageous for the mitigation of effects of the loss-of-coolant accidents (LOCAs) in (1) keeping the containment pressure as low as for the conventional LWRs, (2) injecting water to the containment for cooling the reactor pressure vessel (RPV) and/or flooding a break, and (3) confining the non-condensable gas in the drywell. The gas confinement function makes the system considerably suitable for reactor designs with passive cooling systems utilizing heat exchangers, such as the steam generator (SG) secondary side cooling system for an integral reactor, and the passive containment cooling system (PCCS), because it avoids adverse effects of non-condensable gas on the heat transfer performance during LOCAs. The usefulness of the developed concept is confirmed in the RELAP5/MOD3 code calculation.

JAEA Reports

Chaotic behavior in a system simulating the pressure balanced injection system; Analysis of passive safety reactor behavior, JAERI's nuclear research promotion program, H12-012 (Contract research)

Madarame, Haruki*; Okamoto, Koji*; Tanaka, Gentaro*; Morimoto, Yuichiro*; Sato, Akira*; Kondo, Masaya

JAERI-Tech 2003-017, 156 Pages, 2003/03

JAERI-Tech-2003-017.pdf:5.31MB

no abstracts in English

Journal Articles

Thermal-hydraulic research on future reactor systems in the ROSA program at JAERI

Yonomoto, Taisuke; Otsu, Iwao; Svetlov, S.*

Proceedings of 3rd Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-3), p.521 - 528, 2002/00

A research project is being conducted at the Japan Atomic Energy Research Institute on thermal hydraulics for the future reactor systems. The present paper provides the belief description of the project, followed by two recent topics: the natural circulation in the PWR loop and the condensation heat transfer for a passive cooling system. For the first topic, we discuss the importance of the modeling of the nonuniform flow behavior among SG U-tubes for the assessment of the long-term decay heat removal systems relying on the SG secondary side cooling. Such a system is planned to be used in APWR+, a Japanese next-generation PWR. The condensation heat transfer was investigated using the data obtained at the SPOT test facility in Russia. The results have shown that the measured heat transfer rates on the inner surface of the tube consisting of several bends and short straight sections can be predicted using the existing correlations with the accuracy of several percentage, although the correlations are based typically on the data taken using relatively long straight tube.

Journal Articles

Analysis of multi-dimensional boiling flow in secondary water pool of horizontal PCCS; Effect of pool size

Onuki, Akira; Nakamura, Hideo; Kawamura, Shinichi*; Saishu, Sadanori*

Nippon Kikai Gakkai Netsu Kogaku Koenkai Koen Rombunshu, p.31 - 32, 2001/11

A passive containment cooling system (PCCS) is under planning to use in a next-generation-type BWR for long-term cooling by condensing steam using horizontal heat exchangers. Heat transfer behavior in a secondary water pool is one of important phenomena governing heat removal performance of the PCCS. Boiling and condensation can be supposed under high heat flux regions and the characteristics of the two-phase natural circulation should be evaluated. This study investigated effects of pool size on the characteristics by multi-dimensional two-fluid model code ACE-3D. It was found from the analyses that the pool size gives no significant influences for the characteristics in tube bundle under local-boiling mode.

Journal Articles

Design of advanced integral-type marine reactor, MRX

Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Hoshi, Tsutao*; Sako, Kiyoshi*

Nuclear Engineering and Design, 201(2-3), p.155 - 175, 2000/10

 Times Cited Count:35 Percentile:10.21(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Thermal hydraulic research on next generation PWR's using ROSA/LSTF

Yonomoto, Taisuke; Anoda, Yoshinari

IAEA-TECDOC-1149, p.233 - 246, 2000/05

no abstracts in English

JAEA Reports

Status and subjects of thermal-hydraulic analysis for next-generation LWRs

Subcommittee on Improvement of Reactor Thermal-Hydraulic Analysis Codes

JAERI-Review 2000-002, p.105 - 0, 2000/03

JAERI-Review-2000-002.pdf:6.24MB

no abstracts in English

Journal Articles

Analysis of steam-blanket effect in PCCS with horizontal heat exchanger

Onuki, Akira; Nakamura, Hideo; Anoda, Yoshinari

Dai-7-Kai Doryoku Enerugi Gijutsu Shimpojiumu Koen Rombunshu (00-11), p.258 - 263, 2000/00

no abstracts in English

Journal Articles

ROSA/LSTF experiments on low-pressure natural circulation heat removal for next-generation PWRs

Yonomoto, Taisuke; Otsu, Iwao

Proceedings of 12th Pacific Basin Nuclear Conference (PBNC 2000), Vol.1, p.317 - 329, 2000/00

no abstracts in English

Journal Articles

Safety analysis of a submersible compact reactor SCR for under-sea research vessel

Yoritsune, Tsutomu; Kusunoki, Tsuyoshi; Odano, Naoteru; Ishida, Toshihisa

Proceedings of the 4th JSME-KSME Thermal Engineering Conference, p.1_31 - 1_36, 2000/00

no abstracts in English

JAEA Reports

Journal Articles

Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

Aritomi, Masanori*; Onuki, Akira; Arai, Kenji*; *; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

Nippon Genshiryoku Gakkai-Shi, 41(7), p.738 - 757, 1999/00

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Thermal-hydraulic research on passive safety systems for next-generation PWRs using ROSA/LSTF

Yonomoto, Taisuke

NUREG/CP-0166, 3, p.31 - 52, 1998/00

no abstracts in English

Journal Articles

Nonuniform flow behavior among SG U-tubes occurringduring a low-pressure two-phase natural circulation phase in a PWR

Yonomoto, Taisuke; Anoda, Yoshinari

Nihon Kikai Gakkai Dai-6-Kai Doryoku, Enerugi Gijutsu Shimpojiumu '98 Koen Rombunshu, p.198 - 203, 1998/00

no abstracts in English

Journal Articles

PWR small break Loss-of-Coolant-Accident Experiment at ROSA-V/LSTF with a combination of secondary-side depressurization and gravity-driven safety injection

Yonomoto, Taisuke; ; Kukita, Yutaka

Journal of Nuclear Science and Technology, 34(6), p.571 - 581, 1997/06

 Times Cited Count:5 Percentile:55.72(Nuclear Science & Technology)

no abstracts in English

Journal Articles

ROSA-AP600 Experiment simulating a steam generator tube rupture transient

Nakamura, Hideo; Kukita, Yutaka; L.S.Ghan*; R.R.Schultz*

Proc. of 1997 Int. Meeting on Advanced Reactors Safety, 0, p.1245 - 1252, 1997/06

no abstracts in English

69 (Records 1-20 displayed on this page)