Science and Technology of Nuclear Installations

Two-Phase Flow Heat Transfer in Nuclear Reactor Systems


Publishing date
12 Oct 2012
Status
Published
Submission deadline
25 May 2012

1Reactor Engineering Division, Jožef Stefan Institute, Jamova 39, 1000 Ljubljana, Slovenia

2Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstraße 400, 01328 Dresden, Germany

3Commissariat à l'Énergie Atomique (CEA), 17 rue des Martyrs, 38054 Grenoble, France

4Thermal-Hydraulic Safety Research Division, Korean Atomic Energy Research Institute (KAERI), Daedeok-daero 989-111, Yuseong-gu, Daejeon 305-353, Republic of Korea

5Department of Nuclear Engineering, Texas A&M University, 3133 Tamu, College Station, TX 77843-3133, USA


Two-Phase Flow Heat Transfer in Nuclear Reactor Systems

Description

Heat transfer and phase change phenomena in two-phase flows are often encountered in nuclear reactor systems and are therefore extremely important for their optimal design and safe operation. The complex phenomena observed especially during transient operation of nuclear reactor systems necessitate extensive theoretical and experimental investigations. Researchers and experts are encouraged to submit the research papers summarizing their continuing efforts to tackle various types of heat transfer phenomena in two-phase flows. The contributions dealing with phenomenological understanding, modeling, numerical simulations as well as new experiments and experimental methods are very welcome. Two-phase flow heat transfer related to safety and operation of conventional and advanced designs of nuclear reactor systems and components can be addressed. Potential topics include, but are not limited to:

  • Two-phase flow and heat transfer fundamentals
  • Boiling and condensation heat transfer
  • CHF and post-CHF heat transfer in light water reactors
  • Multiphase heat transfer in advanced reactors
  • Rod-bundle thermal hydraulics
  • Heat transfer phenomena during hypothetical accident conditions in nuclear systems
  • Ex-vessel cooling of reactor pressure vessel
  • Computational methods, modelling, new developments, validation, and application
  • System thermal-hydraulic code analyses of two-phase flow heat transfer
  • CFD analyses or multiscale analyses of two-phase flow heat transfer
  • Experiments and analysis of two-phase flow with heat transfer
  • Advanced instrumentation and measurement techniques

Before submission authors should carefully read over the journal's Author Guidelines, which are located at http://www.hindawi.com/journals/stni/guidelines/. Prospective authors should submit an electronic copy of their complete manuscript through the journal Manuscript Tracking System at http://mts.hindawi.com/ according to the following timetable:


Articles

  • Special Issue
  • - Volume 2013
  • - Article ID 587839
  • - Editorial

Two-Phase Flow Heat Transfer in Nuclear Reactor Systems

Boštjan Končar | Eckhard Krepper | ... | Yassin A. Hassan
  • Special Issue
  • - Volume 2013
  • - Article ID 687494
  • - Research Article

CFD for Subcooled Flow Boiling: Parametric Variations

Roland Rzehak | Eckhard Krepper
  • Special Issue
  • - Volume 2012
  • - Article ID 863190
  • - Research Article

Bubble Departure Diameter Prediction Uncertainty

Marko Matkovič | Boštjan Končar
  • Special Issue
  • - Volume 2012
  • - Article ID 248923
  • - Research Article

CHF Phenomena by Photographic Study of Boiling Behavior due to Transient Heat Inputs

Jongdoc Park | Katsuya Fukuda | Qiusheng Liu
  • Special Issue
  • - Volume 2012
  • - Article ID 214381
  • - Research Article

Aqueous Nanofluid as a Two-Phase Coolant for PWR

Pavel N. Alekseev | Yury M. Semchenkov | Alexander L. Shimkevich
  • Special Issue
  • - Volume 2012
  • - Article ID 951923
  • - Research Article

Depressurization of Vertical Pipe with Temperature Gradient Modeled with WAHA Code

Oriol Costa | Iztok Tiselj | Leon Cizelj
  • Special Issue
  • - Volume 2012
  • - Article ID 247482
  • - Research Article

Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

V. H. Sánchez | M. Thieme | W. Tietsch
  • Special Issue
  • - Volume 2012
  • - Article ID 890815
  • - Research Article

A Derivation of the Nonlocal Volume-Averaged Equations for Two-Phase Flow Transport

Gilberto Espinosa-Paredes
Science and Technology of Nuclear Installations
 Journal metrics
Acceptance rate32%
Submission to final decision60 days
Acceptance to publication39 days
CiteScore1.700
Impact Factor0.825
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