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Journal Articles

Thermal-hydraulic analysis of the LBE spallation target head in JAEA

Wan, T.; Obayashi, Hironari; Sasa, Toshinobu

Nuclear Technology, 205(1-2), p.188 - 199, 2019/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Considerations on phenomena scaling for BEPU

Nakamura, Hideo

Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00

no abstracts in English

Journal Articles

Study on the thermal-hydraulic of TEF-T LBE spallation target in JAEA

Wan, T.; Obayashi, Hironari; Sasa, Toshinobu

Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 13 Pages, 2017/09

JAEA Reports

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

Journal Articles

Thermal mixing characteristics of helium gas in high-temperature gas-cooled reactor, 1; Thermal mixing behavior of helium gas in HTTR

Tochio, Daisuke; Fujimoto, Nozomu

Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03

 Times Cited Count:1 Percentile:14.27(Nuclear Science & Technology)

The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.

Journal Articles

Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor

Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*

Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09

 Times Cited Count:11 Percentile:73.32(Nuclear Science & Technology)

A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.

JAEA Reports

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

JAEA Reports

Nuclear Energy System Department annual report

Department of Nuclear Energy System

JAERI-Review 2003-004, 236 Pages, 2003/03

JAERI-Review-2003-004.pdf:16.34MB

This report summarizes the research and development activities in the Department of Nuclear Energy System during the fiscal year of 2001 (April 1, 2001 - March 31, 2002). The Department has been organized from April 1998. The main research activity is aimed to build the basis of the development of future nuclear energy systems. The research activities of the Department cover basic nuclear data evaluation, conceptual design of a reduced-moderation water reactor, reactor physics experiments and development of the reactor analysis codes, experiment and analysis of thermal-hydrodynamics, energy system analysis and assessment, development of advanced materials for a reactor, lifetime reliability assessment on structural material, development of advanced nuclear fuel, design of a marine reactor and the research for a nuclear ship system. The maintenance and operation of reactor engineering facilities belonging to the Department are undertaken. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report.

JAEA Reports

Nuclear Energy System Department annual report; April 1, 2000 - March 31, 2001

Department of Nuclear Energy System

JAERI-Review 2002-005, 280 Pages, 2002/03

JAERI-Review-2002-005.pdf:18.05MB

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

Feasibility study for improvement of efficient irradiation with LEU core in JMTR

Naka, Michihiro; Nagao, Yoshiharu; Komukai, Bunsaku; Tabata, Toshio

Proceedings of 7th Meeting of the International Group on Research Reactors (IGORR-7) (CD-ROM), 7 Pages, 1999/10

no abstracts in English

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

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

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Concept of passive safety light water reactor system (JPSR)

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Enginering (ICONE), Vol. 2, 0, p.723 - 728, 1995/00

no abstracts in English

Journal Articles

Conceptual design of the JAERI passive safety reactor and its thermal-hydraulic characteristics

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

Transactions of the American Nuclear Society, 71, p.527 - 529, 1995/00

no abstracts in English

Journal Articles

Conceptual design of JAERI passive safety reactor (JPSR) and its thermal-hydraulic characteristics

Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

10th Proc. of Nuclear Thermal Hydraulics, 0, p.3 - 12, 1994/00

no abstracts in English

Journal Articles

Fuel temperature analysis method for channel-blockage accident in HTTR

Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro*; *

Nucl. Eng. Des., 150, p.69 - 80, 1994/00

 Times Cited Count:5 Percentile:48.74(Nuclear Science & Technology)

no abstracts in English

Journal Articles

CHF experiments under steady-state and transient conditions for tight lattice core with non-uniform axial power distribution

Iwamura, Takamichi; Watanabe, Hironori; Okubo, Tsutomu; Araya, Fumimasa; Murao, Yoshio

Journal of Nuclear Science and Technology, 30(5), p.413 - 424, 1993/05

 Times Cited Count:1 Percentile:19.22(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Loss-of-coolant accident analysis for ITER divertor system

Yonomoto, Taisuke; Kukita, Yutaka; Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi; *; Takatsu, Hideyuki

Proc. of the 6th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics,Vol. 2, p.807 - 814, 1993/00

no abstracts in English

Journal Articles

Feasibility study of double-flat-core type HCPWR for reactivity initiated accidents

Araya, Fumimasa; Iwamura, Takamichi; Okubo, Tsutomu; ; Murao, Yoshio

Nihon Genshiryoku Gakkai-Shi, 34(8), p.776 - 786, 1992/08

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

32 (Records 1-20 displayed on this page)