Among Reactivity Initiated Accidents (RIAs) for Pressurized Water Reactor (PWR), Control Element Assembly Ejection (CEAE) accident causes the rapid positive reactivity insertion to the core. It causes an asymmetric power distortion which results in the rising of local fuel temperature, fuel pellet thermal expansion and cladding ballooning or rupture. In the CEAE accident, Doppler feedback has a profound effect because the negative reactivity insertion due to the rise of fuel temperature reduces the core power after rapid power excursion. But the Doppler reactivity can’t be calculated properly in the safety analysis code, using point kinetics model, because the point kinetics model is not able to consider spatial-time effect of the sudden rise in local fuel temperature on Doppler feedback calculation during CEAE accident. And then the excessively high core power which results from the underestimated Doppler feedback would make more severe results such as PCMI fuel failure, fuel cladding rupture and serious DNB fuel failure. Therefore, Doppler Weighting Factor (DWF) is needed for the safety analysis of CEAE accident to compensate a missing spatial-time effect on Doppler feedback calculation. In this study, the adequacy of the application of DWF for APR1400 was evaluated by using nuclear design code called ASTRA (Advanced Static and Transient Reactor Analyzer)[1] and a methodology called ISAM (Integrated Safety Analysis Methodology)[2]. ASTRA is the 3D nuclear design code newly developed by KNF and has various functions such as the static core design, the transient core analysis and the operational support. ISAM is the methodology which is newly developed by KNF to perform the Non-LOCA safety analysis by using RETRAN[3] code which is widely used in the transient analysis and based on the point kinetics model.
Skip Nav Destination
ASME 2013 International Mechanical Engineering Congress and Exposition
November 15–21, 2013
San Diego, California, USA
Conference Sponsors:
- ASME
ISBN:
978-0-7918-5629-1
PROCEEDINGS PAPER
A Study on Doppler Weighting Factor for Control Element Assembly Ejection Accident by Using Newly Developed Nuclear Design Code and Non-LOCA Methodology Available to Purchase
Kyungmin Yoon,
Kyungmin Yoon
KEPCO Nuclear Fuel, Daejeon, Korea
Search for other works by this author on:
Chansu Jang,
Chansu Jang
KEPCO Nuclear Fuel, Daejeon, Korea
Search for other works by this author on:
Jooil Yoon
Jooil Yoon
KEPCO Nuclear Fuel, Daejeon, Korea
Search for other works by this author on:
Kyungmin Yoon
KEPCO Nuclear Fuel, Daejeon, Korea
Chansu Jang
KEPCO Nuclear Fuel, Daejeon, Korea
Jooil Yoon
KEPCO Nuclear Fuel, Daejeon, Korea
Paper No:
IMECE2013-65996, V06BT07A058; 3 pages
Published Online:
April 2, 2014
Citation
Yoon, K, Jang, C, & Yoon, J. "A Study on Doppler Weighting Factor for Control Element Assembly Ejection Accident by Using Newly Developed Nuclear Design Code and Non-LOCA Methodology." Proceedings of the ASME 2013 International Mechanical Engineering Congress and Exposition. Volume 6B: Energy. San Diego, California, USA. November 15–21, 2013. V06BT07A058. ASME. https://doi.org/10.1115/IMECE2013-65996
Download citation file:
9
Views
Related Proceedings Papers
Related Articles
COBRA-TF Simulation of DNB Response During Reactivity-Initiated Accidents Using the NSRR Pulse Irradiation Experiments
ASME J of Nuclear Rad Sci (July,2016)
Combining RAVEN, RELAP5-3D, and PHISICS for Fuel Cycle and Core Design Analysis for New Cladding Criteria
ASME J of Nuclear Rad Sci (April,2017)
An Analysis of the Rupture Behavior of Pressurized Fast Reactor Cladding Tubes Subjected to Thermal Transients
J. Eng. Mater. Technol (July,1979)
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Effect of Chromium Content on the On-Cooling Phase Transformations and Induced Prior-β Zr Mechanical Hardening and Failure Mode (in Relation to Enhanced Accident-Tolerant Fuel Chromium-Coated Zirconium-Based Cladding Behavior upon and after High-Temperature Transients)
Zirconium in the Nuclear Industry: 20th International Symposium
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)