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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Control Engineering of Steam Turbine Valve

B. Halimi Kune Y. Suh

A Homogenization Approach for Studying Two-Phase Heat Exchanger Performance

Pacio, J.C Dorao, C.A

Use of Forward Sensitivity Analysis Method to Improve Code Scaling, Applicability, and Uncertainty (CSAU) Methodology

Haihua Zhao Vincent A. Mousseau

Investigation on Sodium Droplet Impingement Phenomenon during Steam Generator Tube Failure Using MPS Method

Jinbiao XIONG Seiichi KOSHIZUKA Mikio SAKAI Hiroyuki OHSHIMA

Uncertainty Analysis of Molten Corium-Concrete Interaction during a Postulated Degraded Core Accidents for a Typical PWR

Soo Y. Park Kwang I. Ahn

Study on a Nuclear Spaceship for Interplanetary Cruise Reactor Thermal-hydraulics Design by using CFD code

Atsurou YOSHIDA Tadashi NARABAYASHI Masashi TSUJI

On the CSAU Employment during ULOF Accident Initiating Phase for Sodium-Cooled Fast Reactors

M. Pellegrini H. Ninokata H. Endo

Thermal-Hydraulic Analysis of an Intermediate LOCA Test at the ROSA Facility including Uncertainty Evaluation

J. Freixa T-W. Kim A. Manera

OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests (PSBT) - Scoping Calculations with the Advanced Subchannel Code CTF

A. Rubin M. Avramova H. Utsuno

Numerical Simulations on Transient Conjugated Heat Transfer within IFMIF High Flux Test Module

Y. Chen F. Arbeiter V. Heinzel G. Schlindwein

Thermal Hydraulics’ Study in the Original Design of TRIGA 2000 Reactor

Rosalina Fiantini Efrizon Umar Damawidjaya Biksono

Numerical Study on Supercritical Fluids Flow and Heat Transfer under Buoyancy

Zhang Hao Xie Zheng-Rui Yang Yan-Hua

Local Hydrodynamics and Mass-transfer of Coolant in VVER-1000 FA with Mixing Grids

Dmitriev Sergey Borodin Sergey Legchanov Maxim Khrobostov Alexsander Solntsev Dmitry Yuan Ming Li

Peach Bottom Instability Analysis with a RELAP5/PARCSv2.7 Detailed Thermal-hydraulic-Neutronic Model

A. Abarca T. Barrachina R. Miró G. Verdú

Development of Tritium Permeation Analysis Code (TPAC)

Eung S. Kim Chang H. Oh Mike Patterson

Dynamic Methodologies of Digital Instrumentation and Control Systems for Probabilistic Risk Assessments in Nuclear Power Plant

Pengfei Liu Akira Yamaguchi Takashi Takata

Numerical Calculations for Steam-Water CCFL Tests Using the 1/3rd Scale Rectangular Channel Simulating a PWR Hot Leg

Ikuo Kinoshita Michio Murase Yoichi Utanohara Dirk Lucas Christophe Vallée Akio Tomiyama

Study of evaluating transient phenomena in Hierarchical Scaling process

Li Shengqiang Jiang Shengyao

A High Fidelity JTopmeret? Model for Gas-Cooled Reactors

Zen Wang Y.C. Lee Evgeny Grachev Robert Luh Frik Geyser Piet De Villiers

Thermal-Hydraulic Safety Analysis for ANGRA1 Nuclear Power Plant After Replacement Steam Generator, Insertion of High Performance Fuel Assembly and Power Uprate

Sydney da Silva Gomes Franklin Palheiros