Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 31756
The Analysis of TRACE/FRAPTRAN in the Fuel Rods of Maanshan PWR for LBLOCA

Authors: J. R. Wang, W.Y. Li, H.T. Lin, J.H. Yang, C. Shih, S.W. Chen


Fuel rod analysis program transient (FRAPTRAN)  code was used to study the fuel rod performance during a postulated  large break loss of coolant accident (LBLOCA) in Maanshan nuclear  power plant (NPP). Previous transient results from thermal hydraulic  code, TRACE, with the same LBLOCA scenario, were used as input  boundary conditions for FRAPTRAN. The simulation results showed  that the peak cladding temperatures and the fuel centerline  temperatures were all below the 10CFR50.46 LOCA criteria. In  addition, the maximum hoop stress was 18 MPa and the oxide  thickness was 0.003mm for the present simulation cases, which are all  within the safety operation ranges. The present study confirms that this  analysis method, the FRAPTRAN code combined with TRACE, is an  appropriate approach to predict the fuel integrity under LBLOCA with  operational ECCS.



Digital Object Identifier (DOI):

Procedia APA BibTeX Chicago EndNote Harvard JSON MLA RIS XML ISO 690 PDF Downloads 2443


[1] J. H. Yang, J. R.Wang and H. T.Lin, C. Shih, Maanshan PWR nuclear power plant using TRACE”, International Conference on Advances in Energy Engineering (ICAEE) Vol. 14, pp. 292–297, 2012.
[2] U.S. NRC, "TRACE V5.0 user manual
[3] U.S. NRC, "U. S. Code of Federal Regulations, Title 10, Energy, Parts0 to 50,”1997.
[4] K. Geelhood, "Modeling high b Proceedings of 2010 LWR Fuel Performance
[5] A. Daavittila, A. Hämäläinen performance analysis with VTT’s coupled code system and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications, 2005.
[6] K. J. Geelhood, W. G. Luscher, 1.4: a computer code for the transient analysis of oxide fuel rods NUREG/CR-7023, Vol. 1, 2011
[7] K.J. Geelhood, W.G. Luscher, C.E. 1.4: integral assessment,” NUREG/CR
[8] Y. Lee, T.J. Mckrell and M. S. Kazimi carbide and its application to LWR reflood,” Int. Congress on Advances in Nuclear Power Plants (ICAPP ’13), Jeju Island, Korea, April 14
[9] Y. S. Bang, A. J. Cheong thermal-hydraulic response and uncertainty of cal of APR1400,”Int.Congress on Advances in Nuclear Power Plants (ICAPP ’13),Jeju Island, Korea, April 14 2013 Paper No. FA002, 2013.