A new loss of flow transient analysis method for OPR1000 (Optimized Power Reactor 1000, previously called KSNP: Korean Standard Nuclear Power Plant) based on RETRAN code were developed. The reference plant for the analysis is Ulchin Unit 3 and the transient analyzed is 4 pump coastdown. The current analysis for loss of RCS flow transient of OPR1000 uses COAST and CESEC codes. The new method uses RETRAN code to replace COAST and CESEC codes. Since the ability of RETRAN to replace CESEC has been studied in other non-LOCA transients, this paper will focus on COAST code and RCP coastdown flow rates. The results from simplified RETRAN nodalization corresponding to COAST show good agreement with RCS flow results from COAST code. The results are also compared with RETRAN basedeck for safety analysis which is more complex and show similar trends. Therefore, previous analysis method for loss of flow of OPR1000 using COAST code can be replaced with the new analysis method based on RETRAN.
Skip Nav Destination
14th International Conference on Nuclear Engineering
July 17–20, 2006
Miami, Florida, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-4243-6
PROCEEDINGS PAPER
The Development of Loss of Flow Analysis Method for OPR1000 Using RETRAN
Dong Hyuk Lee,
Dong Hyuk Lee
Korea Electric Power Research Institute, Daejeon, Korea
Search for other works by this author on:
Yo Han Kim,
Yo Han Kim
Korea Electric Power Research Institute, Daejeon, Korea
Search for other works by this author on:
Chang Kyung Sung
Chang Kyung Sung
Korea Electric Power Research Institute, Daejeon, Korea
Search for other works by this author on:
Dong Hyuk Lee
Korea Electric Power Research Institute, Daejeon, Korea
Yo Han Kim
Korea Electric Power Research Institute, Daejeon, Korea
Chang Kyung Sung
Korea Electric Power Research Institute, Daejeon, Korea
Paper No:
ICONE14-89387, pp. 385-391; 7 pages
Published Online:
September 17, 2008
Citation
Lee, DH, Kim, YH, & Sung, CK. "The Development of Loss of Flow Analysis Method for OPR1000 Using RETRAN." Proceedings of the 14th International Conference on Nuclear Engineering. Volume 2: Thermal Hydraulics. Miami, Florida, USA. July 17–20, 2006. pp. 385-391. ASME. https://doi.org/10.1115/ICONE14-89387
Download citation file:
5
Views
Related Proceedings Papers
Related Articles
Analysis of the Thermal Hydraulic Consequences Following Common Mode Pump Seizure in a Nuclear Power Plant
J. Pressure Vessel Technol (November,2002)
Transient Analysis of Solid Oxide Fuel Cell Hybrids—Part I: Fuel Cell Models
J. Eng. Gas Turbines Power (April,2006)
FAST Code System: Review of Recent Developments and Near-Future Plans
J. Eng. Gas Turbines Power (October,2010)
Related Chapters
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Maine Yankee Atomic Power Plant
Decommissioning Handbook
Dynamic Behavior of Pumping Systems
Pipeline Pumping and Compression Systems: A Practical Approach