Qualification of Coupled 3-D Neutron-Kinetic/Thermal-Hydraulic Code Systems by the Calculation of Main-Steam-Line-Break Benchmarks in an NPP with VVER-440 Reactor-Reference-Cited by-同舟云学术

Qualification of Coupled 3-D Neutron-Kinetic/Thermal-Hydraulic Code Systems by the Calculation of Main-Steam-Line-Break Benchmarks in an NPP with VVER-440 Reactor

Published:2007-11 Issue:3 Volume:157 Page:280-298
ISSN:0029-5639
Container-title:Nuclear Science and Engineering
language:en
Short-container-title:Nuclear Science and Engineering

Author:

Kliem S.¹,Danilin S.²,Hämäläinen A.³,Hádek J.⁴,Keresztúri A.⁵,Siltanen P.⁶

Affiliation:

1. Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119 D-01314 Dresden, Germany

2. Russian Research Center Kurchatov Institute, Institute of Nuclear Reactors, Kurchatov Sq. 1 123182 Moscow, Russia

3. VTT Technical Research Centre of Finland, P.O.B. 1000, FIN-02044 VTT, Finland

4. Nuclear Research Institute Rez plc., 250 68 Rez, Czech Republic

5. KFKI Atomic Energy Research Institute, P.O.B. 49, H-1525 Budapest 114, Hungary

6. Fortum Nuclear Services, Ltd., P.O.B. 100, FIN-00048 FORTUM, Finland

Publisher

Informa UK Limited

Subject

Nuclear Energy and Engineering

Link

https://www.tandfonline.com/doi/pdf/10.13182/NSE07-A2728

Reference12 articles.

1. Validation of coupled neutron kinetic/thermal–hydraulic codes. Part 1: Analysis of a VVER-1000 transient (Balakovo-4)

2. Validation of coupled neutron kinetic/thermal-hydraulic codes. Part 2: Analysis of a VVER-440 transient (Loviisa-1)

3. Validation of coupled codes using VVER plant measurements

4. Using the OECD/NRC Pressurized Water Reactor Main Steam Line Break Benchmark to Study Current Numerical and Computational Issues of Coupled Calculations

5. OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

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