PREDICTION OF FUEL TEMPERATURE OF AP1000 DUE TO THE FORMATION OF CRUD AND OXIDE LAYER
DOI:
https://doi.org/10.17146/tdm.2017.19.2.3521Keywords:
Fuel temperature, Crud, Oxide layer, COBRA-EN, AP1000Abstract
An analysis to predict the fuel temperature due to crud and oxide layer formed on the hot sub-channel cladding surface of AP1000 reactor has been performed. During reactor operation, the heat transfer and cooling process occur on the fuel cladding surface. During the heat exposure process, an oxide layer and crud are formed on the cladding surface. The decrease of heat transfer performance will increase the fuel and cladding temperatures. Therefore, the effect of fuel temperature increase during the heat exposure process has to be analyzed. The analysis was conducted for nominal power of 3400 MWt using COBRA-EN code, by varying the modular oxide thickness of 0, 20, 40, 60, 80, 100 and 120 mm, crud thickness of 0, 10 and 20 mm and black oxide thickness of 0, 10, 20, 30 and 40 mm. For full cycle hot sub-channel condition, the combination of crud thickness of 20 mm and modular oxide thickness of 115 mm give prediction of the peak fuel center line temperature and the peak cladding surface temperature of 1870.73°C and 609.40°C, respectively. However, the oxide layer is predicted only formed on hot sub-channel during BOC (about 40% of full cycle). The results show that the prediction of the peak fuel center line temperature and the peak cladding surface temperature are 1713.18°C and 451.87°C, respectively. Compared to the normal and fresh fuel conditions, the peak fuel center line temperature and the peak cladding surface temperature increase by 6.53% and 29.86%, respectively.
References
Kim J.S., Kim Y.S. Effect of Thermal History on the Termonal Solid Solubility of Hydrogen in Zircaloy-4. International Journal of Hydrogen Energy 2014; 39: 16422- 16449.
Buongiorno J. Can Corrosion and CRUD actually improve safety margin in LWRs? Annals of Nuclear Energy 2014; 63: 9-21.
Siefken L.J., Coryell E.W., Harvego E.A., Hohorst J.K. MATPRO – A Library of Material Properties for Light Water Reactor Accident Analysis. NUREG/CR-6159. Vol 4. Rev. 2. Chapter 5. 2001. p.1-15.
Lee Y., Lee J.I., No H.C. Impacts of Transients Heat Transfer Modeling on Prediction of Advanced Cladding Fracture During LWR LBLOCA. Nuclear Engineering and Design 2016; 298: 25-32.
Zhang L., Bao Y., Tang R. Selection and corrosion evaluation tests of candidate SCWR fuel cladding materials. Nuclear Engineering and Design 2012; 249: 180-187.
Jin H.J., Kim T.K. Neutron irradiation performance of Zircaloy-4 under research reactor operating conditions. Annals of Nuclear Energy 2015; 75: 309-315.
Lee J.S., Shin A., Kang S.S., Woo S.W. Assessment of Fuel Rod Performance by Consideration of Crud Deposition. Transaction of the Korean Nuclear Society Spring Meeting. Gyeongju, Korea, May 29-30, 2008.p.245-246.
Basile D., Behgi M., Chierici R., Salina E., Brega E., “COBRA-EN : Code System for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores”, RSICC Code Package PSR-507, May, Oak-Ridge (2001)
Isnaini, M.D., Widodo S., Subekti M.. Thermal-Hydraulics Analysis on Radial and Axial Power Fluctuation for AP1000 Reactor. Tri Dasa Mega 2015; 17(2): 79-86.
Isnaini M.D., Mutiara E., A Comparison in Thermal-Hydraulics Analisys of PWR-1000 using fixed and temperature function of thermal conductivity. Jurnal Pengembangan Energi Nuklir 2016;18(1): 31-38.
Isnaini M.D., Subekti M.Validation of SIMBAT-PWR Using Standard Code COBRA-EN on Reactor Transient Condition. Tri Dasa Mega 2016; 18(1): 41-50.
Rahgoshay M., Tilehnoee M.H. Optimizing a gap conductance model applicable to VVER-1000 thermal-hydraulic model. Annals of Nuclear Energy 2012;50: 263-267.
Aghaie M., Zolfaghari A., Minuchehr M., Norouzi A. Enhancement of COBRA-EN capability for VVER reactors calculations. Annals of Nuclear Energy 2012;46: 234-243.
Rahmani Y., Pazirandeh A., Ghofrani M.B., Sadighi M. Calculation of the Deterministic Optimum Loading Pattern of the BUSHEHR VVER-1000 Reactor Using the Weighting Factor Method. Annals of Nuclear Energy 2012;49: 170-181.
Rahimi M.H., Jahanfaria G. Thermal-Hydraulic Core Analysis of the VVER-1000 Reactor Using Porous Media Approach. Journal of Fluids and Structures 2014;51: 85-96.
Zarifi E., Jahanfarnia G., Veysi F., “Sub-channel Analysis of Nanofluids Application to VVER-1000 Reactor. Chemical Engineering Research and Design 2013; 91: 625-632.
Suman S., Khan M.K., Pathak M., Singh R.N., Chakravartty J.K. Rupture behaviour of nuclear fuel cladding during loss of coolant accident. Nuclear Engineering and Design 2016; 307: 319-327.
Pshenichnikov A., Stuckert J., Walter M. Microstructure and Mecahinal Properties of Zircaloy-4 Cladding hydrogenated at temperatures typical for loss-of-coolant accident (LOCA) conditions. Nuclear Engineering and Design 2015; 283: 33-39.
Wu H., Udagawa Y., Narukawa T., Amaya M. Crack formation in cladding under LOCA quench conditions. Nuclear Engineering and Design 2016; 303: 25-30.
Nakamura J. Nuclear Fuel Engineering. Reactor Engineering, International Training Course. Japan Atomic Energy Agency, September 2014.
