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Journal Articles

Proposal of benchmark problem of thermal striping phenomena in planar triple parallel jets tests for fundamental code validation in sodium-cooled fast reactor development

Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08

Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.

Journal Articles

Engineering aspects in modeling of high burnup LWR fuel behavior

Suzuki, Motoe

Proceedings of 2nd Japan-Korea-China (5th Japan-Korea) Seminar on Nuclear Reactor Fuel and Materials, p.4 - 10, 2004/03

In designing a fuel performance code which describes complicated interactions working in high burnup fuel, the code will inevitably become a complex structure of inter-dependent models. In normal operation conditions, PCMI occurs and the pellet-clad firm bonding layer makes the cladding to be subjected to a bi-axial stress state, i.e. under tough mechanical loading. In contrast, the bonding layer enhances thermal conductance, decreases the pellet temperature and keeps the pellet-clad contact, resulting in increased resistance against the Lift-Off. For pellet behaviors, the fission gas bubble growth is strongly dependent on temperature, so that a reliable prediction of fuel temperature is required by pellet radial meshing which can fully accommodate the burning analysis results and the rim structure growth. The presentation deals with modeling method in terms of specific aspects such as meshing.

Journal Articles

Detailed dose assessment for the heavily exposed workers in the Tokai-mura criticality accident

Endo, Akira; Yamaguchi, Yasuhiro; Takahashi, Fumiaki

Radiation Risk Assessment Workshop Proceedings, p.151 - 156, 2003/00

We have developed a new system using numerical simulation technique for analyzing dose distribution in various postures by neutron, photon and electron exposures. The system consists of mathematical human phantoms with movable arms and legs and Monte Carlo codes MCNP and MCNPX. This system was applied to the analysis of dose distribution for the heavily exposed workers in the Tokai-mura criticality accident. The paper describes the simulation technique employed and a summary of the dose analysis.

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:32.66(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Transport calculation codes and cross section libraries

Mori, Takamasa; Kosako, Kazuaki*

Purazuma, Kaku Yugo Gakkai-Shi, 71(12), p.1212 - 1219, 1995/12

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:17.52(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

Improvement of pressure drop caluculation model in TRAC-PF1 code

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

JAERI-Data/Code 94-006, 40 Pages, 1994/07

JAERI-Data-Code-94-006.pdf:1.26MB

no abstracts in English

JAEA Reports

Development of REFLA/TRAC code for engineering work station

Onuki, Akira; ; Murao, Yoshio

JAERI-M 94-026, 60 Pages, 1994/03

JAERI-M-94-026.pdf:1.81MB

no abstracts in English

JAEA Reports

Assessment of TRAC-PF1/MOD1 code for core thermal hydraulic behavior during reflood with CCTF and SCTF data

; Onuki, Akira; *; Murao, Yoshio

JAERI-M 93-032, 190 Pages, 1993/03

JAERI-M-93-032.pdf:3.0MB

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

Development of reflood model for two fluid model code based on physical models used in REFLA code

; Murao, Yoshio

Journal of Nuclear Science and Technology, 29(7), p.642 - 655, 1992/07

no abstracts in English

Journal Articles

Analysis of liquid metal MHD fluid flow and heat transfer using the KAT code

Kunugi, Tomoaki; M.S.Tillack*; M.A.Abdou*

Fusion Technology, 19, p.1000 - 1005, 1991/05

no abstracts in English

JAEA Reports

Development of a one-dimensional atmospheric model(PHYDIV3)

Yamazawa, Hiromi

JAERI-M 90-128, 36 Pages, 1990/08

JAERI-M-90-128.pdf:1.16MB

no abstracts in English

Journal Articles

Numerical simulation of heat transfer in the field of light water reactor safety research

Murao, Yoshio

Dennetsu Kenkyu, 26(101), p.101 - 119, 1987/00

no abstracts in English

JAEA Reports

Analysis of Flow in Three Dimensional Vessel using Thermal-hydrauric Analysis Code "STREAM"

; ; Kaminaga, Masanori; Sudo, Yukio

JAERI-M 86-093, 66 Pages, 1986/07

JAERI-M-86-093.pdf:2.01MB

no abstracts in English

JAEA Reports

Vectorization of Nuclear Codes and Numerical Methods

*; Harada, Hiro;

JAERI-M 85-143, 70 Pages, 1985/09

JAERI-M-85-143.pdf:1.76MB

no abstracts in English

26 (Records 1-20 displayed on this page)