会议专题

The 8th International Topical Meeting on Nuclear Thermal-Hydraulics,Operation and Safety(第八届反应堆热工水力、运行和安全国际会议 NUTHOS-8)

总文献量: 243篇会议类型: 国际会议会议地点: 上海主办单位: 中国核学会会议日期: 2010-10-10

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Numerical Simulation of Free Surface Behavior In Spallation Target With Openfoam

Xiang Chai Guanyu Su Hanyang Gu Xu Cheng

Thermal-Hydraulic System Study of a High Pressure, High Temperature Helium Loop Using RELAP5-3D Code

X. Jin B.E. Ghidersa

FIRE WATER PERFORMANCE AS A SEVERE ACCIDENT MANAGEMENT MEASURE IN THE STATION BLACKOUT ACCIDENT

Te-Chuan Wang Lang-Chen Wang Show-Chyuan Chiang

Droplet-Induced Heat Transfer Augmentation Downstream of a Spacer Grid During Dispersed Flow Film Boiling

F. B. Cheung S.M. Bajorek S. Goverapet Srinivasan

Numerical Simulations of Cap-bubbly Flows Using Fluent

Xia Wang Xiaodong Sun

Thermo-structural Analysis on First Walls Cooling Channel of DEMO Blanket Cooled by Supercritical Water

Chenxiao NI Po HU Xu CHENG

DEVELOPMENT OF A CO2 FREE NUCLEAR TOTAL ENERGY SUPPLY SYSTEM

Tadashi Narabayashi Kotaro Sato Masashi Tsuji Yoichiro Shimazu

Simulation of SBLOCA in RD-14M Test Facility using Computer Code MARS_KS

Joosung Kim Kap Kim Kwangwon Seul

Supercritical Water Heat Transfer Test in A Vertical Tube

Gu H.Y Zhang. G Wen.Q.L Cheng X

Thermophysical Properties of Lithium-Lead Eutectic Alloy for Use in Fusion Reactor Safety Analysis

Ming JIN Yunqing BAI Yican WU

Dynamic Control of Nuclear Electric Propulsion System

B. Halimi Kune Y. Suh

Assessment of the Conservatism in the Evaluation Models of Loss of Coolant Analyses for a Pressurized Water Reactor

M. Lee L. Y. Li K.H. Hsu K.Y. Chiang

CFD Analysis of Core Melt Spreading on the Reactor Cavity Floor

Wan-Sik Yeon Sang-Kun Lee Kwang-Hyun Bang Youngjo Choi Jaegon Lee

An experimental study on the critical heat flux for low pressure of water in a uniformly heated vertical tube

Zhao Erlei Xiong Wanyu Lang Xuemei

COMPASS Code Development: Validation of Multi-physics Analysis using Particle Method for Core Disruptive Accidents in Sodium-cooled Fast Reactors

S. Koshizuka Y. Uehara F. Inoue Y. Nagamine H. Yugo Y. Yamamoto M. Himi E. Hirano S. Shimizu M. Oue K. Morita T. Arima Y. Tobita H. Yamano T. Ito M. Naitoh N. Shirakawa H. Okada

DEVELOPMENT OF LEVEL 2 PSA METHODOLOGY FOR SODIUM-COOLED FAST REACTORS: (2) DEVELOPMENT OF TECHNICAL BASIS IN THE INITIATING PHASE OF UNPROTECTED EVENTS

Ikken Sato Yoshiharu Tobita Hidemasa Yamano

Study on the Performance Behavior of Passive Containment Cooling System in a Natural Circulation BWR during a Postulated Accident Condition

Mukesh Kumar Vikas Jain A.K. Nayak P. K. Vijayan D. Saha R.K. Sinha

RELAP5 Code Validation in the Framework of the LACANES OECD/NEA Benchmark for HLM Innovative Nuclear System

Francesco Saverio Nitti Paride Meloni

A FACILITY FOR STEAM-GENERATORS AUTOMATIC PROTECTION OF NPP WITN FAST NEUTRONS REACTORS

V. M. Poplavskij S. G. Kaliakin A. V. Drobyshev N. S. Grachev

Phenomenal Investigations of the Thermal-hydraulic Responses of Multi-Dimensional Simulation and Modeling in Core and Downcomer of L2-5 Test of LOFT Using RELAP5-3DK

M. Lee K. Y. Chiang